Published Federal Register Notice- Proposed Rule- Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors 10-31-24

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NUCLEAR REGULATORY
COMMISSION
10 CFR Parts 1, 2, 10, 11, 19, 20, 21,
25, 26, 30, 40, 50, 51, 53, 70, 72, 73, 74,
75, 95, 140, 150, 170, and 171
[NRC–2019–0062]
RIN 3150–AK31

Risk-Informed, Technology-Inclusive
Regulatory Framework for Advanced
Reactors
Nuclear Regulatory
Commission.
ACTION: Proposed rule.
AGENCY:

The U.S. Nuclear Regulatory
Commission (NRC) is proposing to
revise the NRC’s regulations by adding
a risk-informed, performance-based, and
technology-inclusive regulatory
framework for commercial nuclear
plants in response to the Nuclear Energy
Innovation and Modernization Act
(NEIMA). The NRC plans to hold a
public meeting to promote full
understanding of the proposed rule and
facilitate public comments.
DATES: Submit comments by December
30, 2024. Comments received after this
date will be considered if it is practical
to do so, but the NRC is able to ensure
consideration only for comments
received before this date.
ADDRESSES: You may submit comments
by any of the following methods
however, the NRC encourages electronic
comment submission through the
Federal rulemaking website:
• Federal Rulemaking website: Go to
https://www.regulations.gov and search
for Docket ID NRC–2019–0062. Address
questions about NRC dockets to Helen
Chang; telephone: 301–415–3228; email:
Helen.Chang@nrc.gov. For technical
questions contact the individuals listed
in the FOR FURTHER INFORMATION
CONTACT section of this document.
• Email comments to:
Rulemaking.Comments@nrc.gov. If you
do not receive an automatic email reply
confirming receipt, then contact us at
301–415–1677.
• Fax comments to: Secretary, U.S.
Nuclear Regulatory Commission at 301–
415–1101.
• Mail comments to: Secretary, U.S.
Nuclear Regulatory Commission,
Washington, DC 20555–0001, ATTN:
Rulemakings and Adjudications Staff.
• Hand deliver comments to: 11555
Rockville Pike, Rockville, Maryland
20852, between 7:30 a.m. and 4:15 p.m.
eastern time, Federal workdays;
telephone: 301–415–1677.
You can read a plain language
description of this proposed rule at

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SUMMARY:

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https://www.regulations.gov/docket/
NRC-2019-0062. For additional
direction on obtaining information and
submitting comments, see ‘‘Obtaining
Information and Submitting Comments’’
in the SUPPLEMENTARY INFORMATION
section of this document.
FOR FURTHER INFORMATION CONTACT:
Robert Beall, Office of Nuclear Material
Safety and Safeguards, telephone: 301–
415–3874; email: Robert.Beall@nrc.gov;
or Anders Gilbertson, Office of Nuclear
Reactor Regulation, telephone: 301–
415–1541; email: Anders.Gilbertson@
nrc.gov. Both are staff of the U.S. NRC,
Washington, DC 20555–0001.
SUPPLEMENTARY INFORMATION:
Executive Summary
A. Need for the Regulatory Action
On January 14, 2019, the President
signed the Nuclear Energy Innovation
and Modernization Act (NEIMA) into
law (Pub. L. 115–439). NEIMA section
103(a)(4) directs the NRC to ‘‘complete
a rulemaking to establish a technologyinclusive, regulatory framework for
optional use by commercial advanced
nuclear reactor applicants for new
reactor license applications.’’ NEIMA
defines a ‘‘technology-inclusive
regulatory framework’’ as one that is
‘‘developed using methods of evaluation
that are flexible and practicable for
application to a variety of reactor
technologies, including, where
appropriate, the use of risk-informed
and performance-based techniques.’’
NEIMA, as further amended by the
Accelerating Deployment of Versatile,
Advanced Nuclear for Clean Energy Act
of 2024 (ADVANCE Act), defines the
term ‘‘advanced nuclear reactor’’ as ‘‘a
nuclear fission reactor or fusion
machine, including a prototype plant (as
defined in sections 50.2 and 52.1 of title
10, Code of Federal Regulations (as in
effect on the date of enactment of
[NEIMA])), with significant
improvements compared to commercial
nuclear reactors under construction as
of the date of enactment of [NEIMA].’’
The NRC initially considered
establishing the scope of proposed part
53, ‘‘Risk-Informed, TechnologyInclusive Regulatory Framework for
Commercial Nuclear Plants,’’ of title 10
of the Code of Federal Regulations (10
CFR) as being for ‘‘advanced nuclear
plants’’ consisting of one or more
‘‘advanced nuclear reactors’’ as defined
in NEIMA. Based on public discussions
on the use of the term, the NRC
determined that the NEIMA definition,
although broad, did not define
‘‘significant improvements’’ with
enough specificity to implement in NRC
regulations. Additionally, a number of

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stakeholders suggested that the
descriptor, ‘‘advanced,’’ implied
enhanced safety, while the NEIMA
definition includes ‘‘significant
improvements’’ in areas other than
safety enhancements. In response to this
feedback, and to be technology
inclusive, the NRC determined that the
broader term ‘‘commercial nuclear
plant’’ would be preferable.
The current application and licensing
requirements in 10 CFR part 50,
‘‘Domestic Licensing of Production and
Utilization Facilities,’’ and 10 CFR part
52, ‘‘Licenses, Certifications, and
Approvals for Nuclear Power Plants,’’
were primarily developed to address
license requests concerning watercooled reactors, and to address
operational requirements for those types
of reactors. This proposed rule responds
to NEIMA by creating an alternative
regulatory framework for licensing
future commercial nuclear plants. The
new alternative requirements and
implementing guidance would adopt
technology-inclusive approaches and
use risk-informed and performancebased techniques to ensure an
equivalent level of safety to that of
operating commercial nuclear plants
while providing flexibility for licensing
and regulating a variety of technologies
and designs for commercial nuclear
reactors.
B. Major Provisions
Major provisions of this proposed
rule, supported by accompanying
guidance, include the following:
• A new alternative technologyinclusive, risk-informed, performancebased framework that includes
requirements for licensing and
regulating nuclear plants during the
various stages of their life cycles.
• A new alternative technologyinclusive, risk-informed, and
performance-based framework in 10
CFR part 26, ‘‘Fitness for Duty
Programs,’’ developed from existing
requirements in subpart K, ‘‘FFD
Programs for Construction,’’ of part 26.
• A new alternative technologyinclusive and performance-based
security framework in 10 CFR part 73,
‘‘Physical Protection of Plants and
Materials,’’ that includes requirements
for protection of licensed activities at
commercial nuclear plants.
C. Costs and Benefits
The NRC prepared a draft regulatory
analysis to determine the expected
quantitative costs and benefits of this
proposed rule and associated guidance
as well as qualitative factors to be
considered in the NRC’s rulemaking
decision. The conclusion from the

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Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules
analysis is that this proposed rule and
associated guidance would result in net
averted costs to the industry and the
NRC ranging from $53.6 million using a
7-percent discount rate to $68.2 million
using a 3-percent discount rate, using an
assumption of one applicant under 10
CFR part 53. As the number of
applicants increases, so do the
estimated averted costs.
The draft regulatory analysis also
considers qualitative factors, such as
greater regulatory stability,
predictability, and clarity to the
licensing process. These benefits would
result from incorporating advances in
probabilistic risk assessment (PRA) and
other risk-informed analyses and
codifying regulatory enhancements that
currently exist in regulatory guides
(RGs). Another qualitative factor is
promoting a performance-based
regulatory framework that specifies
requirements to be met and provides
flexibility to an applicant or licensee
regarding the information or approach
needed to satisfy those requirements.
For more information, please see the
draft regulatory analysis (available in
the NRC’s Agencywide Documents
Access and Management System
(ADAMS) Accession No.
ML21165A112).

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Table of Contents
I. Obtaining Information and Submitting
Comments
A. Obtaining Information
B. Submitting Comments
II. Background
A. NRC Advanced Reactor Readiness
B. Stakeholder Views on Part 53
Preliminary Proposed Rule Language
III. Discussion
A. Objective and Applicability
B. Need for Changes to the Existing
Regulatory Framework
C. 10 CFR Part 53: Framework
IV. Part 53: Framework
Subpart A—General Provisions
A. Discussion of Definitions in Proposed
Part 53
B. Other General Provisions
Subpart B—Technology-Inclusive Safety
Requirements
Subpart C—Design and Analysis
Requirements
Subpart D—Siting Requirements
Subpart E—Construction and
Manufacturing Requirements
Subpart F—Requirements for Operation
Subpart G—Decommissioning
Requirements
Subpart H—Licenses, Certifications, and
Approvals
Subpart I—Maintaining and Revising
Licensing Basis Information
Subpart J—Reporting and Other
Administrative Requirements
Subpart M—Enforcement
V. Changes to Other Parts of 10 CFR Chapter
I

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10 CFR Part 26
A. Introduction
B. Proposed Changes to Part 26, Subparts
A Through E and I
C. Proposed Requirements for Part 26,
Subpart M
D. Proposed Changes to Part 26, Subpart N
E. Proposed Changes to Part 26, Subpart O
10 CFR Part 50
A. Section 50.160: Emergency
Preparedness for Small Modular
Reactors, Non-Light-Water Reactors, and
Non-Power Production or Utilization
Facilities
B. Appendix B to Part 50: Quality
Assurance Criteria for Nuclear Power
Plants and Fuel Reprocessing Plants
10 CFR Part 73
A. Section 73.100: Technology-Inclusive
Requirements for Physical Protection of
Licensed Activities at Commercial
Nuclear Plants Against Radiological
Sabotage
B. Section 73.110: Technology-Inclusive
Requirements for Protection of Digital
Computer and Communication Systems
and Networks
C. Section 73.120: Access Authorization
Program for Commercial Nuclear Plants
VI. Specific Requests for Comments
VII. Section-by-Section Analysis
VIII. Regulatory Flexibility Certification
IX. Regulatory Analysis
X. Backfitting and Issue Finality
XI. Cumulative Effects of Regulation
XII. Plain Writing
XIII. Environmental Assessment and
Proposed Finding of No Significant
Environmental Impact
XIV. Paperwork Reduction Act
XV. Criminal Penalties
XVI. Voluntary Consensus Standards
XVII. Availability of Guidance
XVIII. Public Meeting
XIX. Availability of Documents

I. Obtaining Information and
Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC–2019–
0062 when contacting the NRC about
the availability of information for this
action. You may obtain publicly
available information related to this
action by any of the following methods:
• Federal Rulemaking Website: Go to
https://www.regulations.gov and search
for Docket ID NRC–2019–0062.
• NRC’s Agencywide Documents
Access and Management System
(ADAMS): You may obtain publicly
available documents online in the
ADAMS Public Documents collection at
https://www.nrc.gov/reading-rm/
adams.html. To begin the search, select
‘‘Begin Web-based ADAMS Search.’’ For
problems with ADAMS, please contact
the NRC’s Public Document Room (PDR)
reference staff at 1–800–397–4209, at
301–415–4737, or by email to
PDR.Resource@nrc.gov. For the
convenience of the reader, instructions
about obtaining materials referenced in

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this document are provided in the
‘‘Availability of Documents’’ section.
• NRC’s PDR: The PDR, where you
may examine and order copies of
publicly available documents, is open
by appointment. To make an
appointment to visit the PDR, please
send an email to PDR.Resource@nrc.gov
or call 1–800–397–4209 or 301–415–
4737, between 8 a.m. and 4 p.m. eastern
time, Monday through Friday, except
Federal holidays.
B. Submitting Comments
The NRC encourages electronic
comment submission through the
Federal rulemaking website (https://
www.regulations.gov). Please include
Docket ID NRC–2019–0062 in your
comment submission. To facilitate NRC
review, please distinguish between
comments on the proposed rule and
comments on the proposed guidance.
The NRC cautions you not to include
identifying or contact information that
you do not want to be publicly
disclosed in your comment submission.
The NRC will post all comment
submissions at https://
www.regulations.gov as well as enter the
comment submissions into ADAMS.
The NRC does not routinely edit
comment submissions to remove
identifying or contact information.
If you are requesting or aggregating
comments from other persons for
submission to the NRC, then you should
inform those persons not to include
identifying or contact information that
they do not want to be publicly
disclosed in their comment submission.
Your request should state that the NRC
does not routinely edit comment
submissions to remove such information
before making the comment
submissions available to the public or
entering the comment into ADAMS.
II. Background
A. NRC Advanced Reactor Readiness
In its ‘‘Policy Statement on the
Regulation of Advanced Nuclear Power
Plants,’’ dated July 8, 1986, the
Commission stated that it considered
the term ‘‘advanced’’ to apply to
reactors that are significantly different
from current (i.e., current in 1986)
generation light-water reactors (LWRs)
then under construction or in operation,
and that ‘‘advanced’’ includes reactors
that provide enhanced margins of safety
or utilize simplified inherent or other
innovative means to accomplish their
safety functions. At the time, certain
high temperature gas-cooled reactors,
liquid metal reactors, and LWRs of
innovative design were considered to be
‘‘advanced.’’ The 1986 policy statement

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provided the Commission’s policy
regarding the review of, and desired
characteristics associated with,
advanced reactors. The NRC updated
this statement in the ‘‘Policy Statement
on the Regulation of Advanced
Reactors,’’ dated October 14, 2008
(Advanced Reactor Policy Statement).
The agency has undertaken many
activities related to advanced reactors,
including issuing an advance notice of
proposed rulemaking titled,
‘‘Approaches to Risk-Informed and
Performance-Based Requirements for
Nuclear Power Reactors,’’ dated May 4,
2006 (71 FR 26267). These efforts were
often done in parallel, and sometimes
interwoven, with the NRC’s efforts to
improve risk-informed and
performance-based approaches within
the agency (e.g., the Commission’s
policy statement, ‘‘Use of Probabilistic
Risk Assessment Methods in Nuclear
Regulatory Activities,’’ dated August 16,
1995 (PRA Policy Statement)).
In 2016, the NRC issued ‘‘NRC Vision
and Strategy: Safely Achieving Effective
and Efficient Non-Light-Water Mission
Readiness’’ (Advanced Reactor Vision
and Strategy Document), in response to
increasing interest in advanced reactor
designs. The NRC considered the
Department of Energy’s (DOE’s)
advanced reactor deployment goals in
developing the Advanced Reactor
Vision and Strategy Document. Since
publication of the document, the NRC
continues to manage its activities to
support the DOE’s deployment goals.
The Advanced Reactor Vision and
Strategy Document identified initiating
and developing a new risk-informed and
performance-based regulatory
framework as a possible long-term goal.
However, the NRC staff’s initial efforts
were focused on resolving policy issues
and developing guidance for licensing
non-LWR technologies under the
existing regulatory frameworks (parts 50
and 52). The NRC staff issues annual
Commission papers on the status and
progress of the NRC staff’s activities
related to advanced reactors (e.g.,
SECY–24–0020, ‘‘Advanced Reactor
Program Status,’’ dated February 27,
2024). These Commission papers
provide status updates for advanced
reactor activities undertaken both prior
to and after initiation of this
rulemaking.
In 2017, the NRC staff prioritized
activities to support the development of
technology-inclusive, risk-informed,
and performance-based licensing
approaches that could be implemented
under the existing regulatory framework
in parts 50 and 52. One key element of
these efforts was the Licensing
Modernization Project (LMP), a cost-

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shared initiative led by nuclear utilities
and supported by DOE. The LMP is a
technology-inclusive, risk-informed,
and performance-based methodology
developed for non-LWR designs. The
LMP provides a systematic and
reproducible process for licensing-basis
event (LBE) selection and evaluation;
classification of structures, systems, and
components (SSCs); and assessment of
defense in depth. The LMP refined the
DOE’s Next Generation Nuclear Plant
Program methodologies to reflect
interactions with the NRC, to address
feedback from industry, and to broaden
the scope of the approach to ensure
applicability to various non-LWR
technologies. The LMP activities led to
the publication and submittal of Nuclear
Energy Institute (NEI) 18–04, Revision 1,
‘‘Risk-Informed Performance-Based
Technology Inclusive Guidance for NonLight Water Reactor Licensing Basis
Development,’’ issued August 2019. The
document indicates that controlling the
frequencies and potential consequences
of a wide spectrum of events is the
primary focus of the LMP approach.
The NRC endorsed the principles and
methodology in NEI 18–04, with
clarifications, in RG 1.233, ‘‘Guidance
for a Technology-Inclusive, RiskInformed, and Performance-Based
Methodology to Inform the Licensing
Basis and Content of Applications for
Licenses, Certifications, and Approvals
for Non-Light-Water Reactors.’’ The
NRC staff sought Commission approval
of the use of LMP and NEI–18–04 in
SECY–19–0117, ‘‘Technology-Inclusive,
Risk-Informed, and Performance-Based
Methodology to Inform the Licensing
Basis and Content of Applications for
Licenses, Certifications, and Approvals
for Non-Light-Water Reactors,’’ dated
December 2, 2019. In that paper, the
staff described the relationship between
the LMP and NEI–18–04 and previous
relevant Commission decisions,
including those described in SECY–93–
092, ‘‘Issues Pertaining to the Advanced
Reactor (PRISM, MHTGR, and PIUS)
and CANDU 3 Designs and their
Relationship to Current Regulatory
Requirements,’’ dated April 8, 1993. The
Commission approved the use of the
LMP methodology and NEI–18–04 as a
reasonable approach for establishing key
parts of the licensing basis and content
of applications for licenses,
certifications, and approvals for nonLWRs in Staff Requirements
Memorandum (SRM) SRM–SECY–19–
0117, dated May 26, 2020. Although the
LMP approach is technology- inclusive,
the industry and NRC staff initially
focused the LMP’s applicability on nonLWRs, both for efficiency and to support

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near-term non-LWR applications under
the existing regulatory framework, such
as the Advanced Reactor Demonstration
Projects supported by DOE.
As stated in the part 53 rulemaking
plan, SECY–20–0032, the NRC staff
developed part 53 by building upon
recent and ongoing activities such as the
LMP approach described in SECY–19–
0117. Such an approach supports
implementing the NEIMA requirement
to use, where appropriate, risk-informed
and performance-based techniques, and
it also capitalizes on previous initiatives
by the industry, DOE, and the NRC,
including the LMP. This approach
highlights the role of PRA in riskinformed and performance-based
approaches to identifying enhanced
safety margins that can be used to justify
operational flexibilities. The proposed
framework is largely based on the
methodology described in SECY–19–
0117 and includes a prominent role for
PRA.
As discussed in section II.B,
‘‘Stakeholder Views on Part 53
Preliminary Proposed Rule Language,’’
of this document, the NRC conducted
extensive public outreach on early
versions of the proposed rule text. Early
versions of the draft proposed rule
included two alternative regulatory
frameworks. One framework (called
‘‘Framework A’’) offered a licensing
approach centered largely on risk
analysis and the other framework
(called ‘‘Framework B’’) largely
replicated the existing licensing
approach in parts 50 and 52 but
modified it to be technology neutral. In
its SRM to SECY–23–0021, ‘‘Proposed
Rule: Risk-Informed, TechnologyInclusive Regulatory Framework for
Advanced Reactors (RIN 3150–AK31),’’
the Commission disapproved the
inclusion of Framework B in this
proposed rule and directed the staff to
provide them within one year an
options paper for possible future use of
the Framework B methodology.
B. Stakeholder Views on Part 53
Preliminary Proposed Rule Language
In SRM–SECY–20–0032, the
Commission directed the NRC staff to
prepare and release preliminary
proposed rule language, followed by
public outreach and dialogue, and then
further revise the language until the
NRC staff had established the rudiments
of its proposed rule for Commission
consideration. To implement the
Commission’s direction, the NRC staff
undertook an unprecedented program of
stakeholder engagement, recognizing the
importance of this rulemaking to the
advanced reactor community and

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interested stakeholders from a broad
range of backgrounds and organizations.
On November 6, 2020, the NRC
published a notification in the Federal
Register (85 FR 71002) describing plans
for the periodic release of preliminary
proposed rule language, meetings with
stakeholders, and the ability of
stakeholders to provide input during the
development of this proposed rule.
Sections of the preliminary proposed
rule language were subsequently
released, and the NRC held numerous
public meetings to discuss the
preliminary proposed rule language and
obtain input from stakeholders. On
December 10, 2021, the NRC published
a second notification in the Federal
Register (86 FR 70423) announcing that
the development of the proposed rule
and related interactions with
stakeholders were being extended until
August 31, 2022.
By the close of the public stakeholder
interactions on August 31, 2022, the
NRC staff had held 24 public meetings
since September 2020. The NRC staff
also met with the Advisory Committee
on Reactor Safeguards (ACRS) in 16
public meetings during this period. By
the close of the public engagement
period on the preliminary proposed rule
language, 126 letters were received on
the preliminary proposed rule language.
Of these 126 letters, 21 were from nongovernmental organizations, 31 were
from the public, one was from Congress,
and the remaining 73 letters were from
NRC licensees, the NEI, and other
industry groups. In addition, the ACRS
wrote four interim letter reports to the
Chair on this rulemaking and issued its
final letter report on November 22,
2022. The letters from stakeholders
provided various points of view and
suggestions for clarifications, additions,
and deletions to the preliminary
proposed rule language. Copies of these
letters may be viewed and downloaded
from the Federal rulemaking website
https://www.regulations.gov, under
docket number NRC–2019–0062. The
inputs received were considered in the
development of this proposed rule.
However, as described during the
various public interactions related to
this rulemaking and in supporting
documents, the NRC will not formally
disposition the questions and
suggestions related to the preliminary
proposed rule language as it will for
public comments received following the
publication of this proposed rule.
III. Discussion
A. Objective and Applicability
The NRC is proposing to add a new,
alternative part to its regulations that

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would set out a risk-informed,
technology-inclusive framework for the
licensing and regulation of commercial
nuclear plants. This new approach
would achieve the following: (1)
continue to provide reasonable
assurance of adequate protection of
public health and safety and the
common defense and security; (2)
promote regulatory stability,
predictability, and clarity; (3) reduce
requests for exemptions from the
current requirements in parts 50 and 52;
(4) establish new requirements to
address non-LWR technologies; (5)
recognize technological advancements
in reactor design; and (6) credit the
possible response of some designs of
commercial nuclear plants to postulated
accidents, including slower transient
response times and relatively small and
slow release of fission products. This
proposed rule would add 10 CFR part
53; subpart M, ‘‘Fitness for Duty
Programs for Facilities Licensed Under
10 CFR Part 53,’’ to Part 26; § 73.100,
‘‘Technology-inclusive requirements for
physical protection of licensed activities
at commercial nuclear plants against
radiological sabotage,’’ § 73.110,
‘‘Technology-inclusive requirements for
protection of digital computer and
communication systems and networks,’’
and § 73.120, ‘‘Access authorization
program for commercial nuclear
plants,’’ as well as make conforming
changes throughout 10 CFR chapter I,
‘‘Nuclear Regulatory Commission.’’
B. Need for Changes to the Existing
Regulatory Framework
The NRC has long recognized that the
licensing and regulation of a variety of
nuclear reactor technologies would
present challenges because the existing
regulatory framework has evolved
primarily to address the LWR designs
that compose the current operating fleet
(widely referred to as Generation II
reactors). The NRC has had many
interactions with designers of various
reactor technologies under
development, sometimes collectively
referred to as advanced reactors (widely
referred to as Generation III/III+ (i.e.,
evolutionary light-water) and
Generation IV (i.e., non-light-water)
reactors). The interactions have
informed the development of policies
and guidance to support the potential
licensing of new and different types of
reactor facilities, some of which may not
utilize LWR designs. The NRC issued its
Advanced Reactor Policy Statement to
provide all interested parties, including
the public, with the Commission’s
views concerning the desired
characteristics of advanced reactor
designs. The NRC further described its

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early efforts to establish a technologyinclusive approach to the regulation of
nuclear reactors in the advance notice of
proposed rulemaking published in 2006.
The NRC acknowledged in its ‘‘Report
to Congress: Advanced Reactor
Licensing,’’ issued August 2012, that
while the safety philosophy inherent in
the current regulations applies to all
reactor technologies, the specific and
prescriptive aspects of those regulations
clearly focus on the current fleet of LWR
facilities.
Congress similarly recognized the
potential benefits of developing a
regulatory infrastructure to support the
development and commercialization of
advanced nuclear reactors.
Consequently, Congress passed NEIMA
in late 2018, and the President signed it
into law in January 2019. NEIMA
directed the NRC to undertake a
rulemaking to establish a technologyinclusive regulatory framework for
optional use by applicants for new
commercial advanced nuclear reactor
licenses. In addition, on July 9, 2024,
the President signed into law the
Accelerating Deployment of Versatile,
Advanced Nuclear for Clean Energy Act
of 2024, also referred to as the
ADVANCE Act. The NRC is evaluating
its plans for implementing the
ADVANCE Act, including how its
regulations, as well as the proposed part
53 or future revisions to it, could be
used to address provisions in the
ADVANCE Act. The ADVANCE Act
contains provisions on a variety of
nuclear-related topics, such as micro
reactors, nuclear reactor license
application reviews, and nuclear fuel. In
Section VI, ‘‘Specific Requests for
Comments,’’ the NRC is requesting
public input on how part 53 could be
revised to better enable its potential use
to implement the ADVANCE Act.
The requirements in part 53 would
support a wide variety of potential
commercial nuclear reactor
technologies. As noted in this
discussion, the current regulatory
framework in parts 50 and 52 evolved
in the context of the current operating
reactor fleet dominated by LWRs and as
a result includes provisions specific to
LWR technologies. While the NRC can
license other reactor technologies under
the current framework by using existing
regulatory flexibilities and the
exemption process, there is significant
interest in developing a regulatory
framework that is flexible enough to
accommodate multiple technologies and
robust enough to ensure a level of safety
equivalent to parts 50 and 52, consistent
with the Commission’s Advanced
Reactor Policy Statement. The
Commission reiterated its safety

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expectations for new reactors in the
SRM for SECY–10–0121, ‘‘Modifying
the Risk-Informed Regulatory Guidance
for New Reactors,’’ dated March 2, 2011:
Because new plant designs incorporate
operating experience from current generation
reactors, severe accident research, and risk
insights from design probabilistic risk
assessments, the Commission expects that
the advanced technologies incorporated in
new reactors will result in enhanced margins
of safety. However, the Commission
continues to expect (consistent with the 2008
Advanced Reactor Policy Statement), as a
minimum, at least the same degree of
protection of the public and the environment
that is required for current-generation lightwater reactors. New reactors with these
enhanced margins and safety features should
have greater operational flexibility than
current reactors.

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However, developing a regulatory
framework that can accommodate a
wide range of technologies while
maintaining an acceptable level of safety
presents significant regulatory
challenges. The existing regulations
have been developed over the course of
decades and reflect changes to address
events discovered through operating
experience. In contrast, part 53 is being
developed to accommodate technologies
that, in some cases, lack significant
operating experience. To address these
challenges, the NRC drew on welldeveloped approaches to licensing to
produce a technology-neutral and robust
regulatory framework. The proposed
regulatory framework would use PRAs
to assess risks, help establish technical
requirements, and manage operations.
The framework builds on the LMP,
which is a technology-inclusive
approach to licensing that leverages
insights from a detailed PRA to provide
applicants with significant design and
operation flexibilities.
C. 10 CFR Part 53: Framework
This proposed rule consists of several
major components, including a new part
53, to be added to 10 CFR chapter I,
revisions for part 26, part 50, and part
73, and conforming changes throughout
10 CFR chapter I.
Part 53 is comprised of subparts A
through M. These provisions are
organized to provide high-level
performance criteria and to specify
requirements to demonstrate
compliance with those performance
criteria throughout major stages of the
life cycle of commercial nuclear plants.
This organization reflects a systemsengineering style approach to the
design, licensing, operation, and
ultimately decommissioning of future
commercial nuclear plants. Organizing
requirements in this manner also
supports performance-based

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approaches. Required programs (e.g.,
radiation protection) and monitoring
(e.g., technical specification (TS)
surveillance) during the operations
phase that are similar to those required
by part 50 would complement the
design and analysis requirements in
subpart C. The performance-based
approach proposed in part 53 also
includes regulatory requirements that
would allow applicants to use a flexible
and graded approach to the performance
of safety functions based on the role of
a particular SSC, human action, or
program in limiting the overall risks to
the public below accepted standards
through balanced measures to prevent
and mitigate possible events.
Proposed subpart M of part 26 would
be new and would be largely consistent
with the objective-based fitness for duty
(FFD) requirements in current subpart
K, ‘‘FFD Programs for Construction,’’ of
part 26 supplemented by select
requirements from subparts A through I,
N, and O of part 26. These requirements
are designed to ensure program
effectiveness, maintain protections
afforded to individuals subject to the
FFD program, and align with FFD
program implementation by parts 50
and 52 licensees. The proposed
requirements are not entirely equivalent
because current subpart K of part 26
only applies during construction of the
commercial nuclear plant, whereas
proposed subpart M of part 26 would
apply during construction, operation,
and decommissioning. Furthermore,
proposed subpart M of part 26 would
allow the use of a variety of biological
specimens for drug testing as well as
innovative technologies for drug and
alcohol screening and testing that are
not described or allowed by the
requirements in subparts A through K,
N, and O of part 26, except under
limited conditions.
Proposed revisions to part 73 would
establish a new technology-inclusive
consequence-based approach for a range
of security areas, including physical
security, cybersecurity, and access
authorization (AA) for commercial
nuclear reactors. The NRC used
operating experience to include
additional regulatory flexibility for a
part 53 licensee’s implementation of
security requirements.
In addition, this proposed rule would
make conforming changes throughout
10 CFR chapter I, by adding ‘‘and part
53’’ where appropriate to account for
the addition of the proposed part 53.

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IV. Part 53: Framework
Subpart A—General Provisions
Subpart A would provide the general
provisions applicable to all applicants
and licensees that would be established
in part 53 for the issuance, amendment,
and termination of licenses, permits,
certifications, and approvals for
commercial nuclear plants licensed
under Section 103 of the Atomic Energy
Act of 1954, as amended (the Act) and
title II of the Energy Reorganization Act
of 1974 (88 Stat. 1242). Subpart A
would include purpose, scope,
definitions, written communications,
employee protections, completeness and
accuracy of information, exemptions,
standards for review, jurisdictional
limits, consideration of attacks and
destructive acts by enemies of the
United States, and information
collection requirements.
The requirements in subpart A would
be largely equivalent to the general
requirements in part 50 that are
applicable to all part 50 applicants and
licensees (specifically, §§ 50.1 through
50.13) but would reference the
corresponding regulations in part 53 in
place of references to part 50.
A. Discussion of Definitions in Proposed
Part 53
This proposed rule would include a
definition section in § 53.020. The
definitions of most terms in § 53.020
would be equivalent to the
corresponding terms defined in: (1)
§§ 50.2, 52.1, and other NRC
regulations; (2) NEI 18–04, as endorsed
by RG 1.233; or (3) American Society of
Mechanical Engineers (ASME)/
American Nuclear Society Risk
Assessment Standard (RA–S)-1.4–2021,
as endorsed for trial use by RG 1.247,
‘‘Acceptability of Probabilistic Risk
Assessment Results for Non-Light-Water
Reactor Risk-Informed Activities.’’ This
is intended to provide clarity and
consistency in terminology where
possible and to utilize past and ongoing
NRC initiatives to support the licensing
of new reactors. Specific deviations
from existing definitions are further
explained in the following paragraphs.
Regarding the definition of
‘‘Commercial nuclear plant’’ and
‘‘Commercial nuclear reactor’’ in
proposed § 53.020, as noted previously,
the NRC initially considered
establishing the scope of part 53 as
being for ‘‘advanced nuclear plants.’’
The preliminary proposed rule language
defined ‘‘advanced nuclear plant’’ as ‘‘a
utilization facility consisting of one or
more advanced nuclear reactors’’ as
defined in NEIMA. NEIMA defines the
term ‘‘advanced nuclear reactor’’ as ‘‘a

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nuclear fission reactor or fusion
machine, including a prototype plant (as
defined in sections 50.2 and 52.1 of title
10, Code of Federal Regulations (as in
effect on the date of enactment of this
Act)), with significant improvements
compared to commercial nuclear
reactors under construction as of the
date of enactment of this Act, including
improvements such as—(A) additional
inherent safety features; (B) significantly
lower levelized cost of electricity; (C)
lower waste yields; (D) greater fuel
utilization; (E) enhanced reliability; (F)
increased proliferation resistance; (G)
increased thermal efficiency; or (H)
ability to integrate into electric and
nonelectric applications.’’
Based on public discussions on the
use of the term, the NRC determined
that the NEIMA definition, although
broad, did not define ‘‘significant
improvements’’ with enough specificity
to implement in NRC regulations.
Additionally, a number of stakeholders
suggested that the descriptor,
‘‘advanced,’’ implied enhanced safety,
while the NEIMA definition includes
‘‘significant improvements’’ in areas
other than safety enhancements. In
response to this feedback, and to be
technology inclusive, the NRC
determined that the broader term
‘‘commercial nuclear plant’’ would be
preferable. The NEIMA definition of
advanced nuclear reactor also includes
fusion technologies. Fusion energy
systems have not been included in the
scope of part 53 but are the subject of
a separate rulemaking activity,
‘‘Regulatory Framework for Fusion
Systems.’’ See NRC docket ID NRC–
2023–0017 on the Federal rulemaking
website http://www.regulations.gov.
The NRC proposes to allow use of part
53 by any ‘‘commercial nuclear plant.’’
The use of the term ‘‘plant’’ versus
‘‘reactor,’’ as used in existing
regulations (i.e., § 50.2), recognizes that
co-located support facilities and
radionuclide sources need to be
considered in the licensing of a facility.
The phrase ‘‘commercial purposes,’’ as
used in the definition of ‘‘commercial
nuclear plant,’’ includes purposes such
as providing process heat for a variety
of industrial applications (e.g.,
desalination, oil refining, hydrogen
production). The NRC has not compiled
a complete list of such commercial
purposes. The definition of
‘‘Commercial nuclear plant’’ refers to a
‘‘Commercial nuclear reactor,’’ which is
defined based on the definition of
‘‘Nuclear reactor’’ in § 50.2. However,
the phrase ‘‘in a self-supporting chain
reaction’’ was removed from the
definition to enable applying part 53 to
accelerator driven systems that use

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special nuclear material (SNM) but that
do not involve self-sustaining chain
reactions. Relatedly, ‘‘Utilization
facility’’ is also defined in § 53.020
based on the definition of that term in
§ 50.2 but is also revised to refer to a
‘‘Commercial nuclear plant’’ as defined
in § 53.020.
The NRC proposes to include a
definition of ‘‘Consensus code or
standard’’ in part 53 that is based on the
use of these terms in the National
Technology Transfer and Advancement
Act of 1995 (NTTAA) (Pub. L. 104–113)
and the Office of Management and
Budget (OMB) Circular No. A–119,
‘‘Federal Participation in the
Development and Use of Voluntary
Consensus Standards and in Conformity
Assessment Activities.’’ As required by
NTTAA, the NRC undertakes the
following activities: (i) consults with
voluntary consensus standards bodies;
(ii) participates with voluntary
consensus bodies in the development of
consensus standards; and (iii) uses
consensus standards as a means to carry
out the NRC’s policy objectives. In part
53, the NRC is not proposing to
incorporate by reference specific codes
and standards as is done under the
existing regulations in § 50.55a, ‘‘Codes
and standards,’’ because some codes
and standards are LWR-specific. Part 53
would require that design features must
be designed using generally accepted
consensus codes and standards but
would not incorporate the specific code
or standard into the NRC’s regulations.
During public meetings, significant
discussions with stakeholders indicated
that future reactor designers were
interested in the use of international
consensus standards that have not yet
been endorsed by the NRC. The
definition proposed in part 53 would
allow for the use of international codes
and standards not previously used in
NRC licensing but recognizes that the
use of any consensus code or standard
would ultimately need to be found
acceptable by the NRC, either through
generic efforts to endorse a code or
standard or on an application-specific
basis during an individual licensing
review.
The proposed definition of
‘‘Construction’’ is slightly different than
the definition in § 50.10—it would cover
the same concept but be applied to a
slightly different scope of activities
based on how SSCs are classified under
part 53. In part 53, the definition of
‘‘Construction’’ is based on the
definition in § 50.10 but modified to
apply to safety-related (SR) and nonsafety-related but safety-significant
(NSRSS) SSCs identified by the design
and analysis requirements in subparts B

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and C to ensure the safety criteria are
met.
Section 53.020 would also add
definitions for terms related to event
selection (LBEs, design-basis accidents
(DBAs), anticipated event sequences,
unlikely event sequences, and very
unlikely event sequences); equipment
classifications (SR, NSRSS, and nonsafety-significant SSCs); performance
metrics (e.g., safety criteria and
functional design criteria); and special
treatment.
The regulation would define ‘‘Safety
criteria’’ in terms of the plant-level
performance-based metrics that would
be provided in §§ 53.210 and 53.220.
The term ‘‘Functional design criteria’’
would be defined as metrics for the
performance of specific SSCs that are
determined from the role of the SSC in
meeting the safety criteria. These are
new terms that have not previously been
defined or used in NRC regulation.
The term ‘‘Safety-related SSCs’’
would refer to those SSCs needed to
meet the safety criteria in § 53.210. The
term ‘‘Non-safety-related but safetysignificant SSCs’’ would mean those
SSCs that are not SR because they are
not relied upon to perform any function
necessary to demonstrate compliance
with § 53.210 but warrant special
treatment because they are relied on to
achieve adequate defense in depth or
perform risk-significant functions. The
term ‘‘Special treatment’’ would be
defined as requirements, such as quality
assurance and programmatic controls,
identified for each design feature to
ensure that the safety criteria are
satisfied and the safety functions are
fulfilled. These requirements would also
ensure that SR and NSRSS SSCs will
provide defense in depth, or perform
risk-significant functions, under service
conditions and with SSC reliabilities
that are consistent with the analysis
required in proposed subpart C.
Structures, systems, and components
designated as SR would also contribute
to defense in depth and risk-significant
functions and may warrant special
treatments beyond those defined for the
SR functions needed for compliance
with § 53.210. The term ‘‘Non-safetysignificant SSCs’’ would mean those
SSCs that are not SR or NSRSS.
The terms ‘‘Design-basis accidents,’’
‘‘Anticipated event sequences,’’
‘‘Unlikely event sequences,’’ and ‘‘Very
unlikely event sequences’’ would be
defined to be different types of
‘‘Licensing-basis events’’ and would also
be largely equivalent to the LMP’s
definitions of DBAs, anticipated
operational occurrences (AOOs), designbasis events (DBEs), and beyond-designbasis events, respectively. The term

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‘‘Design-basis accidents’’ would be
defined as postulated event sequences
that are used to set functional design
criteria and performance objectives for
the design of SR SSCs through
deterministic analyses. Design-basis
accidents would be derived from the
unlikely event sequences from the PRA
and then analyzed in a conservative
approach by prescriptively assuming
that only SR SSCs are available to
mitigate postulated accident scenarios.
Within the LMP methodology, event
sequences with mean frequencies of 1 ×
10¥2/plant-year and greater would be
classified as anticipated event
sequences. Within the LMP
methodology, infrequent event
sequences with mean frequencies of 1 ×
10¥4/plant-year to 1 × 10¥2/plant-year
would be classified as unlikely event
sequences. ‘‘Very unlikely event
sequences’’ would be less likely to occur
than unlikely event sequences. Within
the LMP methodology, rare event
sequences with frequencies of 5 × 10¥7/
plant-year to 1 × 10¥4/plant-year would
be classified as very unlikely event
sequences. While the proposed
terminology for these event sequences
would create some differences between
part 53 and the LMP, part 53 would use
new terms for these event sequences
specifically to avoid conflicts with
terms already used within part 50 and
part 52 to represent different concepts.
Further, because some stakeholder
comments demonstrated confusion
related to the history of beyond-designbasis accidents terminology, these
definitions seek to clarify the event
categories in part 53. The sections of
this preamble related to subparts B and
C provide additional discussion of
LBEs.
B. Other General Provisions
Section 53.040 would govern written
communications and how applications
and other required information must be
submitted to the NRC. These
requirements would be equivalent to
those in § 50.4.
Section 53.050 would establish
requirements for enforcement action to
which a licensee, an applicant, or a
licensee’s or applicant’s contractor or
subcontractor, or an employee of any of
them may be subject for engaging in
deliberate misconduct. These
requirements would be equivalent to
those in § 50.5.
Section 53.060 would prohibit
discrimination against an employee of a
holder or applicant for an NRC license,
permit, design certification (DC), or
design approval, or a contractor or
subcontractor of a holder or applicant
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design approval for engaging in certain
protected activities. Section 53.060 also
would prescribe a procedure for seeking
a remedy for employees who believe
they have been discriminated against for
engaging in such protected activities.
These requirements would be
equivalent to those in §§ 50.7 and 52.5.
Section 53.070 would govern the
completeness and accuracy of
information provided to the NRC. These
requirements would be equivalent to
those in §§ 50.9 and 52.6.
Section 53.080 would govern
exemptions from the requirements of
the regulations in part 53. These
requirements would be equivalent to
those in §§ 50.12 and 52.7.
Paragraphs (a) through (d) of § 50.90
would establish requirements for
standards that the NRC would consider
in determining whether a construction
permit (CP), operating license (OL),
early site permit (ESP), combined
license, or manufacturing license (ML)
under part 53 would be issued to an
applicant. These requirements would be
equivalent to those in §§ 50.40, 50.42,
50.43 and 50.22, respectively.
Requirements equivalent to those in
§§ 50.41 and 50.21 would not be
included in part 53 because they apply
to Class 104 licenses, and part 53 would
not apply to those licenses.
Section 53.100 would require that no
license issued under part 53 would
cover activities which are not under or
within the jurisdiction of the United
States. These requirements would be
equivalent to those in § 50.53.
Section 53.110 would state that
licensees and applicants would not be
required to provide design features or
other measures for the specific purpose
of protection against the effects of
attacks and destructive acts by enemies
of the United States directed against the
facility or deployment of weapons
incident to U.S. defense activities.
These requirements would be
equivalent to those in § 50.13.
Section 53.115 would establish
requirements for rights related to SNM.
These requirements would be
equivalent to those in § 50.54(b) and (c).
Section 53.117 would establish
requirements for license suspension and
rights of recapture of the material or
control of the facility in a state of war
or national emergency declared by
Congress. These requirements would be
equivalent to those in § 50.54(d).
Section 53.120 would establish
requirements for information collection
requirements and OMB approval. These
requirements would be equivalent to
those in § 50.8.

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Subpart B—Technology-Inclusive Safety
Requirements
Proposed subpart B, ‘‘TechnologyInclusive Safety Requirements,’’ would
provide technology-inclusive safety
criteria that would serve as performance
standards for the subsequent
performance-based requirements used
throughout part 53. Subsequent subparts
would define how specific activities
during various stages of the life cycle of
a commercial nuclear plant contribute
to satisfying these high-level
performance standards. The
performance standards in subpart B
would also establish a means to
determine appropriate regulatory
controls for SSCs, human actions, and
programs in the following subparts. For
example, the classification of SR SSCs
would be built upon the proposed safety
criteria in § 53.210, ‘‘Safety criteria for
design-basis accidents.’’ The more
detailed requirements for those SSCs
would then be further defined in the
design and analysis requirements in
subpart C, ‘‘Design and Analysis
Requirements.’’ The activities for
manufacturing, constructing, and
maintaining the SR SSCs would be
governed by subpart E, ‘‘Construction
and Manufacturing Requirements,’’ and
subpart F, ‘‘Requirements for
Operation.’’
Requirements for NSRSS SSCs
warranting special treatment would
likewise be determined under § 53.220,
‘‘Safety criteria for licensing-basis
events other than design-basis
accidents,’’ in subpart B and § 53.460,
‘‘Safety categorization and special
treatment,’’ in subpart C. Regulatory
requirements related to the NSRSS SSCs
would be distinguished from the
regulatory requirements for SR SSCs
throughout part 53. Part 53 would afford
more flexibility to applicants and
licensees regarding how NSRSS SSCs
would be used in the design and
maintained during plant operations, as
compared to SR SSCs.
The collective set of performancebased requirements in part 53 would be
sufficient, if met, for the NRC to make
the findings required to grant an
application for a utilization facility
under Section 182 of the Act that the
utilization of SNM will be in accord
with the common defense and security
and will provide adequate protection to
the health and safety of the public. This
construct would be similar to existing
NRC regulations, which the Commission
has said on many occasions do not
specifically define ‘‘adequate
protection.’’ However, compliance with
NRC regulations may be presumed to
assure adequate protection at a

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minimum. The requirements throughout
part 53 that support demonstrating
compliance with § 53.220 would be
similar to current regulations that both
contribute to assuring adequate
protection of public health and safety
and are desirable to promote the
common defense and security or to
protect health or to minimize danger to
life or property under Section 161 of the
Act.
Consistent with historical practice,
Sections 182 and 161 of the Act are
cited as authorizing legislation within
this proposed rule. However, specific
language from the Act would not be
incorporated into the safety objectives
or safety criteria in part 53. This is
because, again consistent with historical
practice, the NRC would not be defining
‘‘adequate protection’’ through the
individual safety requirements in part
53. Rather, part 53 would enable the
NRC to make its required findings under
the Act by providing sufficient
performance standards, safety criteria,
and related requirements on how
applicants must demonstrate
compliance with subpart B and other
subparts.
Section 53.210 would provide safety
criteria for DBAs that would be required
to be identified under § 53.240 and
analyzed under § 53.450(f) in subpart C
of part 53. Subsequent sections in part
53 would require that the SSCs relied
upon to demonstrate compliance with
the criteria in § 53.210 be classified as
SR. The use of SR SSCs and the 25 rem
reference values for potential
radiological consequences would align
with traditional deterministic
approaches for LWRs from §§ 50.34,
52.79, and 100.11 for evaluating the
effectiveness of plant design features
with respect to postulated reactor
accidents. A footnote similar to that
included in § 50.34(a)(1)(ii)(D)(1) and
§ 52.79(a)(1)(vi)(A) would be included
in § 53.210 to explain that the use of the
25 rem value would not be intended to
imply that this number constitutes an
acceptable limit for an emergency dose
to the public under accident conditions.
Rather, this dose value has been set
forth in this proposed section as a
reference value that would be used in
the evaluation of plant design features
with respect to DBAs to verify that the
proposed designs would provide
assurance of low risk of public exposure
to radiation in the event of an accident.
The inclusion of the safety criteria for
DBAs in subpart B would provide a
logical structure supporting the
identification and treatment of SR SSCs
and establishing the corresponding
functional design criteria for those
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Section 53.220 would provide safety
criteria for LBEs other than DBAs that
would be required to be identified
under § 53.240 and analyzed under
§ 53.450(e) in subpart C. Whereas
§ 53.210 and the related requirements
for SR SSCs would provide that a
defined success path exists for DBAs,
the safety criteria for LBEs other than
DBAs would establish the connections
between SSC design, human actions,
and programmatic controls and a
broader set of potential internal and
external hazards. These safety criteria
would also address defense-in-depth
matters such as a balanced
consideration of prevention and
mitigation.
The safety criterion in § 53.220(b)
would include a requirement to use a
comprehensive risk metric or set of
metrics and associated risk performance
objectives against which calculated
values of the risk metrics are compared.
The comprehensive risk metrics or set of
metrics and associated risk performance
objectives would support a
performance-based approach to
developing an appropriate combination
of design features and programmatic
controls to prevent or mitigate LBEs
other than DBAs. The applicant must
propose the comprehensive risk metric
or set of metrics and associated risk
performance objectives, and the
comprehensive risk metric or set of
metrics and associated risk performance
objectives must provide an appropriate
level of safety. Comprehensive risk
metrics should consist of a proposed
plant risk metric or set of proposed risk
metrics that approximate the total,
overall risk from the facility and that
address the range of possible plant
configurations and associated internal
and external hazards to the extent
practicable. The associated risk
performance objectives are
preestablished, indicative values of the
comprehensive risk metrics that are
used as part of risk-informed decisionmaking. The methodology for
developing and using proposed
comprehensive risk metrics and
associated risk performance objectives is
defined by the proposed requirements
for analyses in § 53.450. Therefore, the
application must include a description
of that methodology and, among other
things, should explain the initial
conditions, boundary conditions, and
key assumptions used to develop and
calculate the risk metrics. Screening
tools and bounding or simplified
methods may be used for any mode or
hazard, provided that the applicant
provides an acceptable technical basis.
As with all risk-informed

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methodologies, treatment of
uncertainties must be addressed.
The risk performance objectives
established under this methodology are
likely to involve assessing and averaging
the risks over a period of time (e.g.,
plant year) and would not constitute a
real-time requirement that must be
continuously demonstrated by the
licensee. The use of a comprehensive
risk metric or set of risk metrics and risk
performance objectives that reflect an
average risk to establish performance
goals for SR and NSRSS SSCs is
consistent with current practices that
use other risk assessment techniques to
address short-term plant configurations
during plant maintenance activities.
It is worth noting that the evaluation
of plant risks, as represented by a
comparison of analysis results to
acceptable risk performance objectives
for comprehensive risk metrics, would
be one of several performance standards
used in subpart B. The proposed use of
multiple performance standards,
including deterministic criteria and
defense-in-depth measures, reflects an
integrated decision-making process
similar to that described in RG 1.174,
‘‘An Approach for Using Probabilistic
Risk Assessment in Risk-Informed
Decisions on Plant-Specific Changes to
the Licensing Basis,’’ Revision 3. The
NRC’s approval of using a
comprehensive risk metric or set of
metrics with associated risk
performance objectives is not, by itself,
an indicator of adequate protection.
Rather, the comparison of
comprehensive risk metrics to
associated risk performance objectives
that are acceptable to the NRC is part of
a suite of regulatory requirements that,
when considered holistically, form the
basis for the NRC’s decision-making.
This is analogous to the approach used
for plants licensed under part 50 and
part 52, where no single regulatory
requirement governs whether a plant is
‘‘safe enough.’’
The RG 1.233, ‘‘Guidance for a
Technology-Inclusive, Risk-Informed,
and Performance-Based Methodology to
Inform the Licensing Basis and Content
of Applications for Licenses,
Certifications, and Approvals for NonLight-Water Reactors,’’ describes an
example of an acceptable approach for
identifying and analyzing LBEs under
part 50 and part 52, including the use
of the quantitative health objectives
(QHOs) stated in the NRC’s policy
statement, ‘‘Safety Goals for Nuclear
Power Plant Operation,’’ dated August
4, 1986 (51 FR 28044), as corrected and
republished August 21, 1986 (51 FR
30028) (Safety Goals Policy Statement),
as acceptable performance objectives for

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comprehensive risk metrics. The use of
comprehensive risk metrics, such as the
individual early fatality risk (IEFR) and
the individual latent cancer fatality risk
(ILCFR), and associated risk
performance objectives, such as the
QHOs, from the Safety Goals Policy
Statement, could form the basis for one
approach to meet § 53.220(b). The
requirement for comprehensive risk
metrics, in combination with the other
proposed requirements in subparts B
and C, would bring the approach
endorsed in RG 1.233 for parts 50 and
52 into part 53. Additionally, the use of
comprehensive risk metrics and
associated risk performance objectives
would provide a logical performance
objective to support the risk
management approaches in the various
subparts comprising proposed part 53.
The Commission stated in the
introduction of the Safety Goals Policy
Statement that improvements to thencurrent regulatory practices could lead
to a more coherent and consistent
regulation of nuclear power plants, a
more predictable regulatory process, a
better public understanding of the
regulatory criteria that the NRC applies,
and public confidence in the safety of
operating plants. Accordingly, the
Commission announced the safety goals
with a focus on the risks to the public
from nuclear power plant operation.
Following the issuance of the Safety
Goals Policy Statement, the NRC has
used the comprehensive risk metrics
and performance objectives provided in
the safety goals within the criteria for
many decisions involving safety
judgments during the licensing and
regulation of operating reactors and
proposed nuclear reactor designs.
Consistent with NUREG–0880, the
proposed comprehensive risk metrics
and associated risk performance
objectives required under § 53.220(b)
could be expressed in terms of a
biologically average individual in terms
of age and other risk factors. Although
some comprehensive risk objectives
such as the IEFR and ILCFR are defined
in terms of fatality risks, the
Commission continues to make clear
that no death attributable to nuclear
power plant operation will ever be
‘‘acceptable’’ in the sense that the
Commission would regard it as a routine
or permissible event. Comprehensive
risk metrics and associated risk
performance objectives as used in this
proposed rule would establish
acceptable risks, not acceptable deaths.
Applicants under the proposed part
53 may choose to develop and seek NRC
approval of comprehensive risk metrics
or sets of risk metrics and associated
risk performance objectives beyond

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those discussed above, including the
use of surrogate measures for use in
specific analyses to satisfy the proposed
requirements in § 53.220(b). Such
surrogate measures for comprehensive
risk metrics and associated risk
performance objectives could be used in
a manner similar to the use of core
damage frequency and conditional
containment failure probability for
LWRs within the safety goal evaluation
process in NUREG/BR–0058,
‘‘Regulatory Analysis Guidelines of the
U.S. Nuclear Regulatory Commission,’’
and other assessments of LWRs using
the NRC’s safety goals. The NRC would,
as appropriate, review novel approaches
for comprehensive metrics and
associated risk performance goals
proposed by applicants, industry
organizations, or standard development
organizations and would engage
stakeholders during the development of
the related regulatory guidance or
specific licensing actions.
Section 53.230 would require safety
functions needed to ensure that the
safety criteria under §§ 53.210 and
53.220 can be met if an assumed LBE
were to occur at a commercial nuclear
plant. Section 53.230 would specify that
limiting the release of radioactive
materials from the facility is the primary
safety function, and therefore, limiting
potential offsite consequences (i.e., dose
to a hypothetical individual) would be
used as the primary performance metric
throughout part 53. The additional or
subsidiary safety functions needed to
limit the release of radionuclides may
include, without limitation, controlling
processes related to reactivity, heat
generation, heat removal, and chemical
interactions. This proposed rule
provides flexibility to applicants and
licensees in identifying, implementing,
and maintaining the safety functions
supporting retention of radionuclides
for commercial nuclear plants of varying
sizes and technologies.
Proposed § 53.240 would require
applicants to identify and address LBEs.
LBEs are unplanned events, resulting
from both internal and external hazards,
that are used in the design and analyses
required under part 53 for licensing
commercial nuclear plants. This ensures
estimates of offsite consequences from
analyses performed under proposed
§ 53.450 are below the safety criteria
identified under proposed §§ 53.210 and
53.220 and that SSCs, personnel, and
programs address the safety functions
from proposed § 53.230. Including a
high-level performance requirement
related to the identification and analysis
of LBEs in subpart B would reflect the
historical and continuing importance of
evaluating unplanned events as part of

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the licensing of commercial nuclear
plants. Proposed § 53.240 would require
identification and analysis of LBEs
under § 53.450, which would require a
PRA. Examples of acceptable methods
of using PRAs to identify and assess
LBEs would be the methodology in RG
1.233, as discussed in Draft Regulatory
Guide (DG)–1413, ‘‘TechnologyInclusive Identification of Licensing
Events for Commercial Nuclear Plants.’’
Section 53.250 would establish
defense-in-depth requirements based on
the longstanding philosophy of
providing defense in depth to address
uncertainties about the design,
operation, and performance of
commercial nuclear plants. For
example, parts 50 and 52 address
defense in depth through layered
prescriptive technical requirements
(e.g., fuel performance, cladding
integrity, reactor coolant system
integrity, containment performance) for
LWRs. In contrast, the flexibility
afforded to applicants in how they
propose to demonstrate compliance
with the high-level safety criteria within
part 53 would necessitate this specific
requirement to ensure defense in depth
is provided. The requirements in this
section would state that no single
engineered design feature, human
action, or programmatic control, no
matter how robust, should be
exclusively relied upon to address LBEs
other than DBAs. The phrase
‘‘engineered design feature’’ would not
preclude the possible crediting of
inherent characteristics within the
design and analysis for commercial
nuclear reactors. While defense in depth
would only be assessed for LBEs other
than DBAs, the need to ensure
dedicated success paths for DBAs would
contribute to the overall defense in
depth for each commercial nuclear plant
under part 53.
Section 53.260 would govern normal
operations and would establish a level
of safety based on current requirements
in 10 CFR part 20, ‘‘Standards for
Protection Against Radiation,’’ which
limits doses to members of the public
and dose rates in unrestricted areas.
Section 53.270 would provide for the
protection of plant workers and would
establish a level of safety based on
current requirements in 10 CFR part 20
which limits occupational dose.
Subpart C—Design and Analysis
Requirements
This subpart would provide
requirements for the design of
commercial nuclear plants and the
supporting analyses, including the
analyses of LBEs, to demonstrate that
the performance standards in proposed

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subpart B can be satisfied. The sections
within subpart C would reflect the
overall hierarchy throughout part 53,
which would cover: (1) plant-level
safety criteria (§§ 53.210, 53.220, and
53.470); (2) safety functions (§ 53.230)
needed to demonstrate compliance with
the safety criteria; (3) design features
(§ 53.400), human actions, and
programmatic controls needed to fulfill
the safety functions; and (4) functional
design criteria (§§ 53.410 and 53.420)
that must be defined for each design
feature relied on to demonstrate the
safety criteria (§§ 53.210, 53.220, and
53.470) are met. Subpart C would also
contribute to the logic and structure of
part 53 by distinguishing between SR
SSCs and NSRSS SSCs and licenseecontrolled programs that address LBEs
other than DBAs. Specifically, SR SSCs,
human actions, and programmatic
controls needed to protect against DBAs
are used to satisfy the safety criteria in
§ 53.210. Non-safety-related but safetysignificant SSCs, human actions, and
licensee-controlled programs that
address LBEs other than DBAs generally
contribute to the appropriate measures
considering potential risks to public
health and safety.
Section 53.400 would establish a
requirement that design features be
provided for each commercial nuclear
plant to satisfy the safety criteria and
fulfill safety functions from proposed
subpart B during LBEs. Other sections
in subpart C would, in turn, further
address the necessary capabilities and
reliabilities for SSCs by establishing
functional design criteria, fulfilling
design requirements, performing
analyses of LBEs, performing other
supporting analyses, and categorizing
SSCs based on their roles in preventing
or mitigating LBEs.
Section 53.410 would require that
functional design criteria be defined for
design features relied upon to
demonstrate that the consequences from
DBAs would be below the criteria in
§ 53.210 through analyses performed
under § 53.450(f), which includes
insights from both PRAs and
deterministic analyses. Other sections
within part 53 would establish
appropriate controls on these design
features (e.g., safety classification,
protection from external hazards,
quality assurance, and TS) to ensure the
functional design criteria are satisfied.
The performance requirements for the
SSCs needed to address DBAs and the
corresponding human actions and
programmatic controls would contribute
to ensuring that a commercial nuclear
plant licensed under part 53 would
meet the safety criteria in § 53.210.

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Section 53.415 would require that SR
SSCs be protected against or designed to
withstand the effects of natural
phenomena (e.g., earthquakes,
tornadoes, hurricanes, floods, tsunami,
and seiches) and constructed hazards
(e.g., from dams, transportation routes,
and military or industrial facilities).
Specifically, § 53.415 would require that
SR SSCs remain capable of performing
the safety functions stated in § 53.230
for which they are credited up to the
design-basis external hazard levels as
determined under § 53.510. As used in
§ 53.415 and subpart D of part 53, a
hazard level would refer to such things
as the magnitude and recurrence rate of
an earthquake and the resultant ground
motions, the height of a flood, the force
of hurricane winds, or the
concentrations of chemicals resulting
from a release from a nearby facility.
These requirements would support
either traditional deterministic
approaches for determining and
protecting against external hazards or
probabilistic approaches that are being
developed for seismic and some other
external hazards.
Section 53.420 would require that
functional design criteria be defined for
design features that play a significant
role in demonstrating that the safety
criteria for LBEs other than DBAs are
satisfied. The analyses required for this
demonstration would be described in
proposed § 53.450(e), which would
require that those events be identified
and assessed using a PRA methodology
in combination with other generally
accepted approaches for systematically
evaluating engineered systems. The
SSCs determined to be safety significant
(i.e., either SR or NSRSS) would have
associated special treatment
requirements as specified in § 53.460.
Special treatment would be defined in
subpart A of part 53 and generally refers
to measures (e.g., quality assurance,
testing, monitoring) taken beyond the
procurement and installation of
commercial grade products to provide
confidence that the SSC will comply
with the applicable functional design
criteria. The inclusion of a systematic
approach to identifying the functional
design criteria for SSCs and tailoring the
special treatments to specific LBEs and
safety functions is an important
contributor to satisfy the proposed
safety criteria in subpart B. Therefore,
designers and licensees for commercial
nuclear plants would be provided
flexibility on how LBEs other than
DBAs are either prevented or mitigated
and how the calculated comprehensive
plant risks satisfy the safety criterion
established under § 53.220(b).

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Section 53.425 would establish
requirements for design features and
related functional design criteria
limiting doses to members of the public
during normal operations to satisfy the
criteria in part 20. Section 53.430 would
provide similar requirements for design
features and related functional design
criteria for protection of plant workers
to meet the safety criteria in part 20.
Similar to existing regulations, the NRC
considers that licensees would generally
comply with the requirements of part 20
to keep doses as low as reasonably
achievable by meeting a design objective
of keeping doses to the public from
routine plant effluents less than 10
millirem per year. This goal is similar to
that provided by appendix I to part 50
and would assist designers, applicants,
and licensees in performing the
evaluations of possible reductions in
public dose from routine effluents when
considering costs and other factors. As
emphasized in existing regulations in
part 50, the design objective of keeping
doses to the public from routine plant
effluents less than 10 millirem per year
should not be construed as a radiation
protection standard. The NRC
anticipates that future guidance will
continue to reflect this performance
goal.
The proposed requirements in
§§ 53.425 and 53.430 for design features
and functional design criteria to support
radiation protection activities have
parallels in existing regulations such as
§ 50.34(a) and (b)(3), which require in
part that the means be provided for
meeting the requirements of part 20 and
General Design Criterion 60, 61, 63, and
64 in appendix A to part 50, which
provide radiation protection related
design criteria.
Section 53.440 would address various
design requirements that warrant
specific mention to ensure that the
design features required by § 53.400
comply with the functional design
criteria required by §§ 53.410 and
53.420. These requirements would be
met through design practices,
consideration of testing and operating
experience, and various assessments of
LBEs and other potential challenges to
commercial nuclear plants. Discussions
of some of the key design requirements
included in this section follow.
• § 53.440(a): An essential element to
ensuring a proposed design can comply
with the performance criteria in
proposed part 53 would be that the
abilities of design features to fulfill their
safety functions are demonstrated by a
combination of analyses, test programs,
prototype testing, and operating
experience. This requirement closely
aligns with the language in § 50.43(e)

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and is proposed in part 53 as the same
foundational requirement. In addition,
the proposed § 53.440(a) would require
the design processes for SSCs under this
section to include administrative
procedures for evaluating operating,
design, and construction experience for
considering applicable important
industry experiences in the design of
those SSCs. This proposed requirement
corresponds to the existing requirement
under § 50.34(f)(3)(i) that was developed
in response to the 1979 accident at
Three Mile Island Nuclear Generating
Station.
• § 53.440(b): The design and
licensing of commercial nuclear plants
should use generally accepted
consensus codes and standards. Such
codes and standards ensure sufficient
testing and qualification of materials
and equipment and provide defined
processes, specifications, and
acceptance criteria for use by designers
and suppliers. The NRC would indicate
acceptance of consensus codes and
standards used in the design and
licensing of a specific commercial
nuclear plant either through the NRC’s
generic endorsement of a code or
standard (i.e., through regulatory
guidance), including any limitations or
conditions, that can be referenced
within an application, or through the
review of a referenced code or standard
as part of the review of a specific
application.
• § 53.440(c): The design
requirements in subpart C would
require the materials used for SR and
NSRSS SSCs to be qualified for their
service conditions over the design life of
the SSC.
• § 53.440(d): The requirements in
§ 53.440 would include the need to
consider possible degradation
mechanisms for materials and
equipment to inform both the design
process and the development of
integrity assessment programs to be
executed during plant operations in
accordance with subpart F of part 53.
The inclusion of requirements related to
designing and monitoring for possible
degradation mechanisms reflects
important lessons learned from the
history of LWRs as well as operating
experience with structures and systems
in countless other engineering
endeavors.
• § 53.440(e) and (f): The design
requirements in subpart C would state
specific design requirements similar to
existing requirements in parts 50, 52,
and 73 for protections against fires and
explosions and consideration of safety
and security together in the design
process.

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• § 53.440(g) and (h): Specific design
requirements are proposed to ensure
that commercial nuclear reactors under
part 53 have the capability to achieve
and maintain subcriticality and longterm cooling. The requirements would
be included to address the potential that
some reactor designs may be able to
achieve a stable end state for the
purpose of event analyses but might
need further actions to completely shut
down and service the facility.
• § 53.440(i): The design, analysis,
and development of programmatic
controls under part 53 would consider
the number of reactor units and other
significant inventories of radioactive
materials contributing to the risks to
public health and safety. This would
reflect the definition of ‘‘Commercial
nuclear plant’’ in subpart A and
reinforce that the evaluation of LBEs is
performed on a plant-wide basis. This
aspect of part 53 would be different
from parts 50 and 52, which generally
define safety requirements on the
assumption of events involving only
individual reactor units.
• § 53.440(j): A design requirement is
proposed to provide a technologyinclusive requirement that would be
equivalent to the requirements in
§ 50.150 to address the possible impact
of a large commercial aircraft.
• § 53.440(k): The inclusion of a
specific proposed requirement to
address the risks to public health from
potential chemical hazards of licensed
material is appropriate given the
diversity of reactor technologies and
designs that might be licensed under
part 53. The requirement in part 53
would be similar to the existing
requirements in 10 CFR part 70,
‘‘Domestic Licensing of Special Nuclear
Material,’’ that address both potential
radiological and chemical hazards for
licensed materials at fuel cycle facilities.
• § 53.440(l): Provisions are proposed
to require that measures be taken during
the design of commercial nuclear plants
to minimize contamination of the
facility and the environment, facilitate
eventual decommissioning, and
minimize the generation of radioactive
waste in accordance with § 20.1406.
• § 53.440(m): A design requirement
is proposed to provide a technologyinclusive equivalent to the requirements
in § 50.68 by including options for
commercial nuclear plants to either
have a monitoring system capable of
detecting a criticality as described in
§ 70.24 or to have restrictions on SNM
handling and storage that would prevent
inadvertent criticality events.
• § 53.440(n): The design would need
to reflect state-of-the-art human factors
principles for safe and reliable

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performance in all settings that human
activities are expected for performing or
supporting the continued availability of
plant safety or emergency response
functions.
Section 53.450 would establish
analysis requirements and would center
upon the use of a PRA in combination
with other generally accepted
approaches for systematically evaluating
engineered systems. The reliance on
PRAs as a key component in the
proposed analysis requirements for part
53 would reflect the decades of
improvements in PRA methodologies
and the increasing use of PRA
techniques in the design, licensing, and
oversight of both operating and future
nuclear reactors. Part of the
Commission’s PRA Policy Statement is
that the use of PRA technology should
be increased in all regulatory matters to
the extent supported by the state of the
art in PRA methods and data and in a
manner that complements the NRC’s
deterministic approach and supports the
NRC’s traditional defense-in-depth
philosophy. The need to supplement
PRA insights with other engineering
approaches and judgments reflects the
NRC’s longstanding policy described in
the SRM to SECY–98–144, ‘‘Staff
Requirements—SECY–98–144—White
Paper on Risk-Informed and
Performance-Based Regulations,’’ dated
February 24, 1999, for regulatory
decision-making to be risk-informed but
not solely based on numerical results of
a risk assessment (i.e., not a risk-based
approach). Part 53 would maintain a
role for NRC’s traditional deterministic
approaches (particularly for DBAs) and
defense-in-depth philosophy by
including specific requirements
utilizing these regulatory tools in
subparts B and C.
PRA would be used in combination
with other techniques in part 53 to
identify and categorize LBEs, classify
SSCs, and evaluate defense in depth.
This increased role for the PRA
necessitates that it would be developed,
performed, and maintained in
accordance with NRC-approved
standards and practices (see § 53.450(c)
and (d)). The computer codes used to
model the plant response and the
behavior of the barriers to the release of
radionuclides would need to be
qualified for the range of conditions
being simulated across a wide range of
unplanned events. These analyses
would need to use realistic approaches
and address uncertainties associated
with states of knowledge, modeling, and
performance of SSCs.
While industry consensus PRA
standards and peer review processes
endorsed in RGs 1.200 and 1.247 remain

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acceptable for developing a PRA, they
are not regulatory requirements and an
application under part 53 need not
follow every aspect of the applicable
consensus PRA standard. Existing
processes for defining the scope and
capability of a PRA supporting an
application offer flexibility in
determining the degree to which the
PRA needs to be developed and may be
informed by other factors such as design
complexity and the needed degree of
realism and level of detail, consistent
with the use of the PRA and substance
of the application. Such processes are
currently available for appropriately
defining the scope of the PRA and
determining applicability of supporting
requirements in consensus PRA
standards needed to satisfy the
proposed regulatory requirements for
the specific uses of analyses under
§ 53.450(b). Likewise, NRC
determinations of the acceptability of
such PRAs would include consideration
of the appropriateness of the applicantdefined scope as part of determining the
applicability of and conformance to
consensus PRA standard supporting
requirements consistent with the
current state of practice. In addition,
these determinations would include
consideration of other aspects of the
development of the PRA, such as PRA
peer reviews. An NRC determination of
the acceptability of a PRA includes but
is not limited to assessing the initial and
boundary conditions and key
assumptions used in the analysis,
treatment of uncertainties, and the use
of screening tools and bounding or
simplified methods for any mode or
hazard, provided the use of those tools
and methods is justified by an
acceptable technical basis. In that
regard, the consensus PRA standards
would not be applied by the NRC as a
strict checklist of requirements for part
53 PRA acceptability determinations.
The proposed § 53.450(c) would
require periodic maintenance and
upgrading of the PRA to maintain an
alignment between the supporting
analyses and the design and
performance of plant equipment,
programs and procedures, and other
factors associated with meeting the
safety criteria of the proposed § 53.220
and the evaluation criteria of proposed
§ 53.450(e)(2). The periodic
maintenance of the PRA would also be
a means to consider new or revised
information related to external hazards,
industry operating experience,
performance issues with or degradation
of SSCs, and other contributors to the
frequency and potential consequences
of various event sequences. The

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periodic assessments performed by
licensees to support the maintenance of
the PRA and other requirements in the
proposed part 53 would be
complemented by NRC inspections and
programs to assess new or revised
information related to topics such as
natural hazards, operating experience,
and potential generic safety issues.
The categories of LBEs used in part 53
would include anticipated event
sequences, unlikely event sequences,
and very unlikely event sequences. The
unlikely event sequences would include
those events with estimated frequencies
well below the frequency of events
expected to occur during the lifetime of
a commercial nuclear plant. An
important aspect of the analysis
requirements is that, under proposed
§ 53.450(e), the analyses of LBEs other
than DBAs would not only be used to
show the performance criteria of
§ 53.220 are satisfied but to also show
that evaluation criteria defined for each
LBE or category of LBEs would also be
satisfied. Such evaluation criteria for
specific LBEs or categories of LBEs
would be defined in terms of limits on
the release of radionuclides or
maintaining the integrity of one or more
barriers used to limit the release of
radionuclides and reflect a graded
approach of allowing lesser potential
consequences from more frequent
events. An example of such evaluation
criteria for a range of LBEs that could
likely be expanded for part 53 is
provided in RG 1.233. Another
proposed requirement for the proposed
§ 53.450(e) analyses is that the
methodology would need to include a
means to identify event sequences
deemed risk-significant such that those
event sequences can be given special
attention within other sections of part
53.
Part 53 would maintain an important
role for a deterministic analysis of DBAs
in the performance criteria of § 53.210
and the related analytical requirements
in § 53.450(f). The analysis of DBAs
would be required to address event
sequences drawn from those with
estimated frequencies below the
expected lifetime of a generation of
reactors (e.g., event sequences with
frequencies as low as one in ten
thousand years). As proposed in this
section, DBAs would need to be
analyzed using deterministic methods
and ensure a safe, stable end state with
reliance upon only SR SSCs and human
actions, if needed, to be performed by
operators licensed under the provisions
of §§ 53.760 through 53.795.
While the DBAs analyzed under part
53 would be similar to the traditional
DBAs analyzed under parts 50 and 52,

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there are important distinctions between
the overall role of DBA analyses in part
50 and proposed part 53. In part 53, the
role of the DBA analysis would be more
narrowly focused on selecting SR SSCs
and determining functional design
criteria for those SSCs to ensure the
commercial nuclear plant meets the
safety criteria in § 53.210. The overall
control of risks posed by commercial
nuclear plants under part 53 would be
provided by the analyses of and
measures taken for both DBAs and other
LBEs, including very unlikely event
sequences. This would contrast with the
traditional deterministic approach in
part 50 wherein the analyses of DBEs
such as DBAs were used to provide
bounding assessments, incorporate
standard design rules such as
assumptions related to single failures,
and to define conservative performance
requirements for SR SSCs. Limitations
related to the traditional deterministic
approach were addressed in part 50
through case-by-case assessments and
specific actions for beyond-design-basis
events such as anticipated transients
without scram and station blackout.
Section 53.450 would also include
provisions to ensure that analyses are
performed to support the design
requirements of § 53.440(e) on fire
protection, § 53.440(j) on aircraft impact
assessments, and § 53.425 on using
design features and plant programs to
control doses to members of the public
from routine effluents and direct
radiation from contained sources. The
proposed analysis requirements related
to fire protection would support either
a traditional, deterministic approach or
a more risk-informed approach where
the risks from fires are addressed within
the identification and analyses of LBEs.
Section 53.460 would establish
criteria for the safety classification of
SSCs and determination of appropriate
special treatments. As noted in subpart
A, the term ‘‘Special treatments’’ would
be defined to mean those items, such as
measures taken to satisfy functional
design criteria, quality assurance, and
programmatic controls, which provide
assurance that certain SSCs will provide
defense in depth or perform risksignificant functions. These
requirements would also provide
confidence that the SSCs will perform
under the service conditions and with
the reliability credited in the analysis
performed in accordance with § 53.450
to satisfy the safety criteria in §§ 53.210
and 53.220. The terminology used in
part 53 would include the following
categories for SSC classification: (1) SR;
(2) NSRSS; and (3) non-safety
significant. Requirements for SR SSCs
would be defined in other sections of

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part 53 and would include using TSs for
controls during operation and the
application of quality assurance
requirements from appendix B of part
50.
Requirements for NSRSS SSCs would
include the need to identify necessary
special treatments such as performance
measures on reliability. Licensees
would generally be afforded flexibility
in maintaining and changing special
treatments for SSCs categorized as
NSRSS. Non-safety-significant SSCs
would be addressed under normal
licensee programs for commercial grade
equipment and typical industry
practices for general plant design and
maintenance. Safety-related SSCs would
also contribute to defense in depth and
risk-significant functions and may
warrant special treatments beyond those
defined for their SR functions to reflect
their role in meeting the safety criteria
in § 53.220 and the evaluation criteria in
§ 53.450(e).
Section 53.470 would allow an
applicant or licensee to seek operational
flexibilities by adopting more restrictive
criteria than those provided in § 53.220
and that might otherwise be used in the
analysis of LBEs under § 53.450(e). Such
an approach might be taken to ensure
sufficient safety margins to gain
operational flexibilities in areas such as
justifying siting in relation to
population centers or staffing levels. As
an example, an applicant or licensee
could propose to justify siting proposals
by adopting alternate criteria for very
unlikely event sequences. Such
alternate criteria could require
calculated consequences for an
individual at the exclusion area
boundary to be less than one rem total
effective dose equivalent (TEDE). This
section would establish requirements to
ensure that, if more restrictive
evaluation criteria than those required
by a methodology were used to justify
operational flexibilities, then the
analysis, design features, and
programmatic controls would be
established and maintained accordingly.
Section 53.480 would establish
seismic design considerations. This
proposed section would relate to the
safety criteria in subpart B, the
analytical requirements related to
external hazards in § 53.450, and
subpart D, ‘‘Siting Requirements.’’ For
licenses issued under part 53, this
section in subpart C would support a
variety of approaches to seismic design.
For example, a design for a commercial
nuclear plant could show that SSCs are
able to withstand the effects of
earthquakes by adopting an approach
similar to that in appendix S to part 50.
Alternatively, an applicant could follow

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the more recent risk-informed
alternatives afforded by standards
development organizations (e.g.,
American Society of Civil Engineers
(ASCE)/Structural Engineering Institute
(SEI) 43–19, ‘‘Seismic Design Criteria for
Structures, Systems, and Components in
Nuclear Facilities.’’) Because the agency
has not endorsed ASCE/SEI–43–19, an
applicant can propose to use ASCE/SEI
43–19 on an application specific basis to
meet § 53.480 and the NRC would
evaluate the adequacy of the standard as
applied in that application. The design
could also be done with the full
integration of seismic PRAs into the
design and licensing of a particular
commercial nuclear plant. This section
has been developed to accommodate a
variety of potential risk-informed,
performance-based seismic design
approaches. The analyses required by
§ 53.450 would need to address seismic
hazards as well as other external
hazards. The expected responses of
SSCs to a range of seismic events would
be included in the analyses when
ensuring that the safety criteria defined
under § 53.220 would be met. The
potential SSC responses to seismic
hazards could be addressed in the
analyses using a fragility model
(conditional probability of its failure at
a given hazard input level), a high
confidence of low probability of failure
value, or other method endorsed or
otherwise found acceptable by the NRC.
Subpart D—Siting Requirements
Proposed subpart D in part 53 would
state requirements for the siting of
commercial nuclear plants and would
serve the role provided by 10 CFR part
100, ‘‘Reactor Site Criteria,’’ for nuclear
reactors licensed under parts 50 and 52.
As reflected in proposed § 53.500, the
reason for establishing siting
requirements would remain the same as
it has been historically, which is to
ensure that licensees and applicants
assess what impact the site environs
may have on a commercial nuclear plant
(e.g., external hazards) and, conversely,
what potential adverse health and safety
impacts a commercial nuclear plant may
have on nearby populations in view of
the site characteristics.
Proposed § 53.510 would require that
design-basis external hazard levels be
identified and characterized based on
site-specific assessments of natural and
constructed hazards with the potential
to adversely affect plant functions. The
site-specific assessments would be used
in the proposed § 53.415, which would
require that SR SSCs be designed to
withstand the effects of natural
phenomena and constructed hazards of
levels or severities up to design-basis

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external hazard levels. The design-basis
levels for external hazards relevant to a
site would need to account for
uncertainties and variabilities in data,
models, and methods used to
characterize those hazards. Existing
approaches could be used to
demonstrate compliance with this
requirement. The historical importance
of assessing seismic events as risks to
commercial nuclear plants and the
associated development of riskinformed approaches to address seismic
events would be reflected in proposed
§ 53.480, ‘‘Earthquake engineering,’’ and
specific requirements in subpart C. The
NRC is developing a graded approach
for seismic design by grouping SSCs
into different seismic design categories
(SDCs) based on their risk significance.
While the agency has not endorsed
ASCE/SEI–43–19, an applicant can
propose to use ASCE/SEI 43–19 on an
application-specific basis to meet
§ 53.480 and the NRC will evaluate the
adequacy of the standard as applied in
that application. The NRC staff will
continue to review ASCE/SEI–43–19 as
part of its efforts to further develop
guidance in this area. The approach
described in RG 1.208, ‘‘A PerformanceBased Approach to Define the SiteSpecific Earthquake Ground Motion,’’
would be an acceptable way to develop
site-specific ground motion response
spectra for SSCs under appendix S to
part 50, which corresponds to SSCs that
are categorized as the highest SDC
(SDC–5) in ASCE/SEI 43–19.
The evaluation of seismic hazards
under subpart D would need to be
sufficient to inform a site-specific
design (e.g., a CP or custom COL) or
confirm the use of a standard design for
a commercial nuclear plant under
§ 53.480 and other sections of subpart C.
A risk-informed approach could use
several design-basis ground motions
(DBGMs) to assess SSCs in various SDCs
(i.e., one DBGM per SDC). Section
53.510(d) would state that geologic and
seismic siting factors must also include
related hazards such as seismically
induced flooding and volcanic activity
that may affect the design and operation
of a proposed commercial nuclear plant
for the proposed site.
Section 53.520 would require
applicants to identify and assess site
characteristics related to topics which
might include meteorology, geology,
hydrology, or other areas in the design
and analyses required under subpart C.
Proposed section 53.530 would set
requirements for population-related
considerations and maintain
requirements and definitions similar to
those currently in part 100 for an
exclusion area, low population zone,

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and population center distance. The
NRC recognizes that some applicants
may propose to essentially collapse the
exclusion area and low population zone
to the site boundary. This approach
would rest on a demonstration that the
calculated consequences of DBAs
remain below the proposed dose
guidelines used in § 53.210, which are
the same as those in the existing
regulations in parts 50, 52, and 100. The
proposed definitions in § 53.020 would
allow such configurations, assuming
they were justified by the design and
analyses from subpart C. This approach
should provide flexibility to justify
alternative exclusion areas and low
population zones without foreclosing
the option for an applicant to define
more conventional exclusion areas and
low population zones outside of a
defined site boundary. The NRC’s longstanding preference for siting reactors in
areas of low population density would
be maintained in part 53 by using the
current language from part 100 in
proposed § 53.530(c). The NRC revised
guidance related to population densities
surrounding a commercial nuclear plant
in Revision 4 to RG 4.7, ‘‘General Site
Suitability Criteria for Nuclear Power
Stations’’ to reflect Commission
direction in SRM–SECY–20–0045,
‘‘Population Related Siting
Considerations for Advanced Reactors.’’
Site-related requirements in part 20
(restricted area) and part 73 (protected
and owner-controlled areas) would
remain applicable to commercial
nuclear plants licensed under part 53.
Proposed section 53.540 would
require that site characteristics be
appropriately considered in other
activities such as the design and
analysis performed under proposed
subpart D and the emergency planning
and security programs under proposed
subpart F.
Subpart E—Construction and
Manufacturing Requirements
The proposed part 53 language would
establish construction and
manufacturing requirements in subpart
E. The proposed language for
construction-related activities would
largely reflect current requirements in
part 50 without any fundamental
changes. Limited changes would be
made in several places, as described in
the following paragraphs, to be
technology-neutral and for consistency
with the organization and language of
part 53. The proposed language for
requirements for manufacturing
activities would largely mirror those for
construction-related activities. However,
the proposed manufacturing
requirements have been updated from

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the current requirements in subpart F of
part 52 to better accommodate the
possible factory fabrication of
manufactured reactors. The
manufacturing of specific components
outside the scope of an ML would not
be addressed by these proposed
subparts.
Section 53.600 would establish the
overall construction and manufacturing
requirements for CPs, OLs, COLs, MLs,
and limited work authorizations
(LWAs). This section would connect the
construction and manufacturing
requirements to the safety criteria,
quality assurance requirements, and
other requirements located in other
subparts. These requirements would
require that construction and
manufacturing activities be managed
and conducted such that when
combined with associated design
features and programmatic controls, the
constructed plant would satisfy the
relevant requirements in subpart B.
Section 53.605 would establish
requirements for the reporting of defects
and instances of noncompliance during
construction. This section would
provide equivalent requirements to
those in § 50.55(e).
Section 53.610(a) would establish the
requirement to have in place a welldefined command and control structure
to manage construction activities. The
requirements would generally reflect
current requirements, with an emphasis
on the quality assurance programs for
complying with the requirements in
appendix B to part 50. The proposed
§ 53.610(a)(6) would require
programmatic controls for implementing
special treatment for NSRSS SSCs to
align with requirements in other
subparts in part 53. The section would
also refer to other NRC regulations to
address matters such as requirements to
have a FFD program, a radiation
protection program if radioactive
materials are brought onto the site, and
security programs to protect sensitive
information and protect against cyber
threats.
Section 53.610(b) would provide
requirements governing construction
activities, including the equivalent of
the requirement in § 50.10(e) that
prohibits starting construction until the
NRC has authorized the activities by
issuing a CP, COL, ESP, or LWA.
Section 53.610(b)(1)(iii) would require
procedures to be in place prior to
beginning construction to ensure that
construction-related activities do not
undermine important features such as
slope stability and that constructionrelated activities such as backfilling of
excavated portions of the site
appropriately address potential pre-

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construction activities such as the
emplacement of retaining walls or
drainage systems. Other requirements in
these paragraphs would be equivalent to
requirements in parts 50 and 52 with
appropriate references to other parts for
items such as possession of byproduct
material or SNM, protecting operating
units from construction activities for
commercial nuclear plants with
multiple reactor units, and having a
redress plan in case LWA activities are
terminated.
Section 53.610(c) would address
inspection and acceptance activities by
including requirements in part 53
equivalent to specific quality assurance
criteria in appendix B to part 50 and
inspections, tests, analyses, and
acceptance criteria (ITAAC) in part 52
for COLs.
Section 53.620(a) would include
proposed requirements covering the
activities performed under an ML issued
under part 53. Provisions related to MLs
were first adopted by the NRC in 1973
through the addition of appendix M to
part 50. The regulation supported the
manufacture of a nuclear power reactor
to be incorporated into a commercial
nuclear plant under a CP and operated
under an OL at a different location from
the place of manufacture.1 The
regulations and processes for MLs were
changed substantially in the part 52
rulemaking in 2007 (72 FR 49352). The
most important shift in the ML concept
in that rulemaking was that a final
reactor design, which would be
equivalent to that required for a
standard DC under part 52 or an OL
under part 50, must be submitted and
approved before issuance of an ML. The
rationale for that change was that
approval of a final design ensures early
consideration and resolution of
technical matters before there is any
substantial commitment of resources
associated with the actual manufacture
of the reactor, which greatly enhances
regulatory stability and predictability.
The proposed part 53 sections in
subpart E for manufacturing and in
subpart H for licensing matters would
maintain requirements equivalent to
those in part 52 for MLs. The NRC
approval of a standard design and
related manufacturing processes,
coupled with a stable workforce and
established procedures, has the
potential for maintaining and even
improving the quality and consistency
of manufacturing, as compared to the
traditional method of constructing
1 On December 17, 1982, the NRC issued
‘‘Manufacturing License ML–1 to Offshore Power
Systems for the manufacture of a maximum of eight
floating nuclear plants,’’ dated September 30, 1982,
but the project was subsequently canceled.

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reactors onsite by a variety of
contractors and subcontractors.
Subpart E would include
requirements that would apply to
portions of a manufactured reactor in
recognition that some activities covered
by an ML may occur at different
fabrication facilities. As with the
preceding sections on construction,
§ 53.620 would establish the
requirements to have in place programs,
procedures, and a well-defined
command and control structure to
manage manufacturing-related
activities.
Section 53.620(b) in subpart E would
propose requirements for executing the
manufacturing activities following
receipt of an ML under part 53.
Information about the design and
manufacturing processes should be
provided by the applicant. The
importance of the ML is reflected in
several of the proposed requirements in
§ 53.620(b) that would refer to
complying with the ML, including
conducting manufacturing processes
within facilities for which the license
holder can control activities. The
essential role of post-manufacturing
inspections would also be incorporated
into this proposed section by requiring
the holder of the ML to perform
inspections and have acceptance
processes for manufactured reactors or
portions of a manufactured reactor.
Section 53.620(c) would provide
proposed requirements for the control of
radioactive materials if the holder of an
ML plans to possess and use source,
byproduct, or SNM as part of the
manufacturing process. By and large,
the proposed subpart E would refer to
NRC regulations in 10 CFR part 30,
‘‘Rules of General Applicability to
Domestic Licensing of Byproduct
Material,’’ 10 CFR part 40, ‘‘Domestic
Licensing of Source Material,’’ and part
70 for the requirements on controlling
radioactive materials. Several specific
requirements to address the potential
hazards of radioactive materials are
proposed in areas such as having a fire
protection program, an emergency plan,
training programs, and procedures to
minimize contamination.
The most significant change proposed
for MLs in part 53 as compared to MLs
under part 52 relates to § 53.620(d) in
subpart E and the associated licensing
provisions in subpart H. These
provisions would allow and establish
requirements for the loading of fuel into
a manufactured reactor at the
manufacturing site for subsequent
transport to a commercial nuclear
facility that will operate pursuant to a
COL. The first requirement in the
proposed § 53.620(d) would establish

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limitations on when a license under part
70 would authorize the loading of fuel
into a reactor manufactured under an
ML. The proposed regulation would
require the manufactured reactor to
include at least two independent
physical mechanisms that will each
prevent criticality should conditions
most favorable to critical operation be
introduced (e.g., optimum neutron
moderation and reflection). This
requirement would contribute to the
NRC’s longstanding practice of requiring
defense in depth for preventing
accidents in any facility dealing with
SNM, including requirements in § 70.64
for certain part 70 licensees to adhere to
the ‘‘double contingency principle.’’
The requirements to have in place
mechanisms to prevent criticality could
likewise support meeting other
provisions in subpart H to part 70, such
as those related to having a safety
program and integrated safety
assessment. The mechanisms to
preclude criticality in the proposed
requirements would reasonably ensure
that a manufactured reactor would not
become critical assuming optimum
neutron moderation, and optimum
neutron reflection conditions. With the
proposed requirements for mechanisms
to prevent criticality and all criticality
safety controls required by 10 CFR part
70 in place, the presence of fuel in the
manufactured reactor would not create
a nuclear hazard different than the
hazard from the presence of the same
fuel in a storage location or container
licensed under 10 CFR part 70.
Collectively, the proposed measures
would reasonably ensure that the
manufactured reactor would not be
capable of operations, thereby obviating
the need for a COL under §§ 53.1416
and 53.1440 to authorize fuel loading.
Additionally, this approach would focus
the ML application and its review on
the design, manufacture, and
deployment of the manufactured
reactor.
The activities involving SNM within
the manufacturing facility, including the
loading of fuel, would be regulated
primarily under the part 70 license. The
reference to the requirements in subpart
H of part 70 in section 53.620(d) assures
that the activities involving the receipt,
storage, and loading of a variety of
possible fuel forms and enrichments at
the manufacturing facility will be
analyzed in a systematic manner and
appropriate protection will be provided
against equipment malfunctions, human
errors, external hazards, and other
adverse conditions. The regulations in
part 51 provide a flexible approach for
environmental review to address the
range of regulated activities under part

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70. The flexibility in part 51 will enable
the NRC to determine the appropriate
type of environmental review based on
the circumstances associated with the
loading of fuel into a specific
manufactured reactor.
The proposed § 53.620(d) cites the
requirements in parts 70, 71, and 73 to
ensure important features and programs
are in place prior to the receipt of SNM.
The features and programs required to
be in place prior to receipt of SNM
include (1) radiation monitoring
instrumentation and alarms; (2)
measures to detect potential criticality
accidents; (3) appropriate procedures,
equipment, and personnel qualified for
the fuel loading; (4) programs for
physical security and cybersecurity; and
(5) material control and accounting
(MC&A) programs. Section
53.620(d)(2)(i) proposes requirements to
address security programs for any ML
authorizing possession of a
manufactured reactor into which fuel
has been loaded at the manufacturing
facility. Currently, for category II SNM,
security measures may be required in
addition to requirements included in
§ 73.67, ‘‘Licensee fixed site and intransit requirements for the physical
protection of special nuclear material of
moderate and low strategic
significance,’’ on a case-by-case basis.
Including appropriate security measures
in the proposed part 53 regulations will
provide additional openness and
transparency for applicants applying for
an ML who seek to load fuel into
manufactured reactors at a
manufacturing site.
Currently, § 73.67 only requires a
security plan for licensees who possess,
use, transport, or deliver to a carrier for
transport SNM of moderate strategic
significance, or 10 kg or more of SNM
of low strategic significance. However,
the proposed physical security program
for fueled manufactured reactors would
require a security plan for any ML
authorizing possession of a
manufactured reactor into which fuel
has been loaded at the manufacturing
facility, regardless of fuel type,
enrichment, and quantity. This is
consistent with other controls for MLs,
including reactivity and criticality
controls.
The proposed requirements would
also require a holder of an ML and part
70 license to address cybersecurity to
ensure a cyberattack would not
adversely impact the functions
performed by digital assets used by the
licensee for physical security, radiation
monitoring, or criticality prevention.
The proposed regulations in part 53
covering the activities related to the
storage, movement, and loading of fresh

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fuel into a manufactured reactor in the
manufacturing facility would likewise
refer to the applicable regulations in
part 70. The proposed § 53.620(d) would
also require the loading or unloading of
unirradiated fuel into or from a
manufactured reactor and any changes
to the configuration of reactivity-related
systems to be performed by a certified
fuel handler meeting the requirements
in subpart F. The NRC is aware of
proposals to introduce reprocessing of
existing or future spent nuclear fuel into
the fuel cycle for some potential
commercial nuclear plants. This
proposed rule does not address the
loading of spent nuclear fuel or fuel
resulting from reprocessing of spent
nuclear fuel into a manufactured
reactor.
Section 53.620(e) would limit the
transport and delivery of a
manufactured reactor or portions of a
manufactured reactor only to a site for
which the Commission has issued a
COL authorizing the construction of a
commercial nuclear plant using a
manufactured reactor under the specific
ML. This proposed requirement is
similar to the limitations in § 52.153,
with the difference being that part 53
would allow the installation of a
manufactured reactor at the site of a
COL but would not include provisions
for installation at a site under a CP. The
possible combination of a manufactured
reactor and the licensing option of CP
and OL seems unlikely and would
require the introduction of ITAAC into
the licensing provisions for a CP and
OL. An additional proposed paragraph
in § 53.620(e) would provide
requirements for protecting fueled
manufactured reactors during transport
to the site of the commercial nuclear
plant by referencing the transportation
and security requirements in 10 CFR
part 71, ‘‘Packaging and Transportation
of Radioactive Material,’’ and part 73.
Section 53.620(f) would include
proposed requirements for the
acceptance and installation of a
manufactured reactor at the site of a
commercial nuclear plant. The proposed
requirements would reference the
construction requirements in § 53.610 to
govern the integration of the
manufactured reactor into the
construction of a commercial nuclear
plant. Other proposed requirements in
the section would address required
receipt inspections and verification that
interface requirements between the
manufactured reactor and the balance of
the commercial nuclear plant have been
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Subpart F—Requirements for Operation
Proposed subpart F would provide the
requirements for the operations phase of
a commercial nuclear plant to ensure
that the safety criteria in subpart B are
satisfied throughout the plant’s lifetime
and during all modes of normal
operation and unplanned events.
Section 53.700 would provide the
overall objectives and general
organization of subpart F, which would
be to establish requirements during
operations for: (1) plant SSCs; (2) plant
personnel; and (3) plant programs.
Proposed § 53.710 would provide the
requirements for maintaining
capabilities, availability, and reliability
of SSCs to demonstrate compliance with
the safety criteria and design
requirements for unplanned events that
are described in proposed subparts B
and C. The basic structure of this
proposed section would be that controls
for SR SSCs are provided by TS and
controls for NSRSS SSCs are required to
be addressed with licensee-controlled
documents and procedures.
The general content and control of TS
under the proposed part 53 would be
similar to the requirements in part 50.
The proposed requirements for TS
would include limits on the inventories
of radioactive materials, plant operating
limits, and specific requirements for
each SR SSC, including limiting
conditions for operation (LCO) and
required surveillances. The proposed
requirements for TS would also include
a section on important design elements,
which is similar to design features in
§ 50.36, and a section for administrative
controls. A provision addressing the
development and submittal of TS to
address decommissioning activities
would also be included in the proposed
subpart G.
The proposed requirements for TS
under part 53 would not carry over
safety limits or associated limiting
safety system settings from § 50.36,
which contains TS requirements for
operating reactors under parts 50 and
52. As discussed in SECY–18–0096,
systematic assessments and more
mechanistic approaches to evaluating
source terms support an alternative
approach to establishing barrier-based
safety limits. An example provided in
that paper is a comparison of: (1) the
traditional specified acceptable fuel
design limits (SAFDL) that support
protecting a specific barrier from
potential failure mechanisms (e.g.,
departure from nucleate boiling to
protect fuel cladding); and (2) the
specified acceptable system
radionuclide release design limit
(SARRDL) concept, which limits the

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possible increase in circulating
radionuclide inventory during normal
operations or an AOO as part of an
integrated or ‘‘functional containment’’
approach. Additional discussion of the
use of SARRDL in the design and
licensing of advanced reactors is
provided in RG 1.232. The SARRDL
could be addressed as an operating limit
within this proposed construct of
requirements for TS. In cases, such as
LWRs, where a SAFDL approach might
be used as part of a mechanistic
approach to meeting the design and
analysis requirements in subpart C, the
associated functional design criteria
proposed in § 53.410 and TS under the
proposed § 53.710(a) would define
similar requirements as those provided
by the safety limit and limiting safety
system setting requirements in § 50.36.
The proposed requirements for TS
under part 53 would not include
specific criteria for identifying when
LCOs must be established (i.e., would
not include an equivalent to
§ 50.36(c)(2)(ii)). Instead, consistent
with subparts B and C, the TS
requirements in subpart F of part 53
would define TS LCOs as providing
limits on SR SSCs. The SR SSCs protect
against DBAs to demonstrate
compliance with the safety criteria in
the proposed § 53.210. In the proposed
construct for part 53, risk-significant
SSCs would be addressed through a
combination of TS for the SR SSCs and
establishment and monitoring of
performance standards for NSRSS SSCs.
In addition to addressing TS for SR
SSCs, proposed § 53.710 would require
appropriate controls be developed and
implemented for NSRSS SSCs.
Examples include appropriate
surveillances and controls established
through reliability assurance programs.
Configuration management and other
special treatments would provide that
the capabilities, availabilities, and
reliabilities of NSRSS SSCs are
maintained consistent with the
underlying risk assessments while
providing flexibility to licensees
through maintaining the management
functions within licensee-controlled
programs. Controls on NSRSS SSCs are
appropriate as part of the overall
performance-based approach within
proposed part 53. Special treatments
beyond those defined for their SR
functions may also be warranted for SR
SSCs to reflect their role in meeting the
safety criteria in § 53.220 and the
evaluation criteria in § 53.450(e). The
performance objectives for NSRSS SSCs
would reflect that the comprehensive
risk metrics and related risk
performance objectives established
under § 53.220 may involve assessing

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and averaging the risks over a defined
period (e.g., plant year) and would not
constitute a real-time requirement that
must be continuously demonstrated by
the licensee. The controls under
§ 53.710(b) justify proposed changes in
part 53 from the traditional or
deterministic approaches in parts 50
and 52 in areas such as replacing the
single-failure criterion with a
probabilistic reliability criterion (see
SRM–SECY–03–0047, ‘‘Policy Issues
Related to Licensing Non-Light-Water
Reactor Designs,’’ dated June 26, 2003).
This approach could also support the
incorporation of risk insights and
analytical margins to gain operational
flexibilities in areas such as siting and
staffing requirements described in
subsequent sections of proposed subpart
F.
Proposed § 53.715 would provide the
requirements for developing and
implementing a program to do the
following: (1) control maintenance
activities; (2) take appropriate corrective
action when performance issues are
identified; (3) conduct routine
evaluations of effectiveness; and (4)
assess and manage risks resulting from
maintenance activities. These proposed
requirements are similar to those
included in § 50.65 (maintenance rule),
including the need to assess and manage
the increase in risk that may result from
the proposed maintenance activities.
While, for the maintenance rule,
specific criteria must be developed to
capture both SR and non-SR but
otherwise important SSCs, the proposed
§ 53.715 would cover SR SSCs and
NSRSS consistent with other subparts in
part 53.
Proposed § 53.720 would provide the
requirements for responding to a
seismic event during the operating
phase of the life cycle of a commercial
nuclear plant and would be equivalent
to the requirements in paragraph
IV(a)(3) of appendix S, ‘‘Earthquake
Engineering Criteria for Nuclear Power
Plants,’’ to part 50.
The proposed part 53 would include
provisions to address staffing, training,
personnel qualifications, and human
factors engineering (HFE) in a manner
that is risk informed, technology
inclusive, performance based, and
flexible in nature. During the
development of part 53, the staff
prepared a draft white paper on ‘‘Risk
Informed and Performance Based
Human-System Considerations for
Advanced Reactors,’’ to support
interactions with stakeholders and the
ACRS. Key considerations include the
recognition that staffing, operator
qualifications, and HFE are
interconnected areas that must be

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approached in an integrated manner
and, furthermore, that safety functions,
including the means by which they are
fulfilled, provide an effective method
for informing technology-inclusive
requirements.
The requirements associated with this
approach would be in §§ 53.725 through
53.830. Section 53.725 discusses
applicability and defines specific terms.
Some definitions draw from those in
§ 55.4. Several new definitions would be
introduced for use within the context of
subpart F. These new definitions would
be the following: ‘‘Automation,’’
‘‘Auxiliary operator,’’ ‘‘Generally
licensed reactor operator,’’ ‘‘Interactiondependent-mitigation facility,’’ ‘‘Load
following,’’ ‘‘Self-reliant-mitigation
facility.’’
Sections 53.725 through 53.830 would
be divided into four portions that would
cover general operational requirements,
operator and senior operator licensing
requirements, generally licensed reactor
operator (GLRO) requirements, and
general training requirements for plant
staff. The NRC intends to provide
guidance addressing the review of
operator staffing plans; the review of
operator, senior operator, and GLRO
examination programs; and the
implementation of scalable HFE
reviews. Licensees would be required to
use GLROs upon demonstrating
compliance with the criteria in § 53.800.
Certain routine communications are
necessary to facilitate the operator
licensing process. The NRC is proposing
to adapt the requirements of §§ 55.5 and
50.74 to § 53.726 to accomplish this.
Specific information must be
collected in order to facilitate the initial
issuance of operator licenses, as well as
to allow for license renewals and
required updates thereafter. Such
information collection activities must
also be approved by the OMB. The NRC
is proposing to adapt the requirements
of § 55.8, to include any needed updates
in OMB approval information, to
§ 53.120 to accomplish this.
The information used within the
regulatory processes of the NRC must be
free from omissions and inaccuracies to
facilitate effective regulation. Consistent
with this, the NRC is proposing to adapt
the requirements of § 55.9 to § 53.728 to
require the completeness and accuracy
of material information provided by
individual applicants and license
holders.
Section 53.730 would provide
performance-based and technologyinclusive requirements for assessing the
role of personnel in facility safety,
applying human-system considerations
within facility design, and incorporating
operational approaches that are

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consistent with design-specific safety
considerations. Most of these
requirements would be adapted from
portions of §§ 50.34(f) and 50.54 and 10
CFR part 55, ‘‘Operators’ Licenses,’’
with considerable modification in order
to reflect the introduction of new
technologies and possible changes in
the roles of personnel in preventing and
mitigating events. The NRC is proposing
that these technical requirements
would, together, serve as a component
of the required content of applications
for OLs and COLs under part 53.
Additionally, the NRC proposes that the
specific technical requirements
associated with HFE, human-system
interface design, concept of operations,
functional requirements analysis, and
function allocation would serve as a
component of the required content of
applications for standard DCs, standard
design approvals, MLs, and CPs, as well.
Human factors engineering is
essential to facilitate the role of
personnel in facility safety in a manner
that is both effective and reliable. The
NRC proposes to adapt § 53.730(a) from
the HFE design requirements of
§ 50.34(f)(2)(iii). A key difference would
be that the requirement would now be
focused on settings where personnel
fulfill their safety or emergency
response roles wherever they may
occur. The NRC additionally proposes
to include within the scope of this
requirement activities for assuring the
continued availability of plant
equipment that is needed for safety, and
envisions that this may encompass
relevant maintenance, inspections, and
testing as well. The NRC intends that
this requirement would be associated
with staff guidance for conducting
scalable reviews of HFE that is planned
to accompany part 53.
Human-system interfaces provide
vital information to operators across a
spectrum of operating conditions that
can range from normal operations
through severe accident conditions. The
specific types of information that must
be available to support operations staff
during such conditions include, in part,
those associated with safety function
parameters, safety system status,
possible core damage states, barrier
integrity, and radioactive leakage. Due
to the importance of such information,
the NRC proposes under § 53.730(b) to
require such human-system interface
design features for all facilities,
irrespective of other flexibilities
proposed under part 53. Therefore, the
NRC proposes to adapt specific postThree Mile Island requirements of
§ 50.34(f) in a technology-inclusive
manner as detailed in the following:

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• Paragraph (b)(1) would be adapted
from § 50.34(f)(2)(iv).
• Paragraph (b)(2) would be adapted
from § 50.34(f)(2)(v).
• Paragraph (b)(3) would be adapted
from § 50.34(f)(2)(xi), 50.34(f)(2)(xii),
and 50.34(f)(2)(xxi).
• Paragraph (b)(4) would be adapted
from § 50.34(f)(2)(xvii), 50.34(f)(2)(xviii),
50.34(f)(2)(xix), and 50.34(f)(2)(xxiv).
• Paragraph (b)(5) would be adapted
from § 50.34(f)(2)(xxvi).
• Paragraph (b)(6) would be adapted
from § 50.34(f)(2)(xxvii).
In addition to the requirements of
§ 53.730(b)(1) through (6), a further set
of human-system interface design
requirements applicable only to those
facilities that will be staffed by GLROs
would be provided under § 53.730(b)(7).
This prescriptive set of design
requirements for those facilities which
demonstrate compliance with the
criteria of § 53.800 would recognize that
the application of HFE under § 53.730(a)
is anticipated to be significantly
reduced at such facilities in the absence
of an expected operator role for the
fulfillment of safety functions. However,
it should be noted that the capability for
an immediately initiated, manual
reactor shutdown would be
conservatively mandated irrespective of
any other design considerations.
The NRC proposes § 53.730(c) to
require the submittal of a concept of
operations that is of sufficient scope and
detail to appropriately inform the staff.
The development of a concept of
operations can facilitate a clear
understanding on the part of the NRC
for potential novel operating concepts.
Additionally, such information is likely
to reduce the degree of resources and
interactions needed for the NRC to
obtain the understanding necessary to
enable flexible requirements in areas
such as staffing, operator qualifications,
and HFE.
The NRC proposes § 53.730(d) to
require the submittal of both a
Functional Requirements Analysis and a
Function Allocation. The identification
of design-specific safety functions and
how they are fulfilled serves as a
primary means for achieving
technology-inclusive requirements
within areas such as staffing, operator
qualifications, and HFE. The Functional
Requirements Analysis and Function
Allocation processes (which are both
HFE methods derived from systems
engineering principles), provide an
effective means to identify both how
safety functions will be satisfied and
how to characterize any associated
operator role in doing so. A Functional
Requirements Analysis shows what
features, systems, and human actions

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are relied upon to demonstrate safety
(i.e., fulfill safety functions). A Function
Allocation then describes how safety
functions are assigned to both personnel
and automatic systems. However, an
important adaptation of the Function
Allocation for use under the proposed
rule would be the further need to not
only describe allocations of safety
functions to human action and
automation, but also to identify
allocations made to active safety
features, passive safety features, or
inherent safety characteristics as well.
Operating experience provides an
important source of information by
which to inform various aspects of
facility design and operations.
Accordingly, the NRC proposes in
§ 53.730(e) to adapt the requirements of
§ 50.34(f)(3)(i) for requiring an operating
experience program.
New technologies may involve
concepts of operations that are more
conducive to customizable licensed
operator staffing requirements than the
prescriptive requirements of § 50.54(m).
Analyses and assessments that are based
on HFE principles provide a
performance-based means of
determining licensed operator and
senior operator staffing needed to
support safe operations. In contrast, for
those facilities required to be staffed by
GLROs, the NRC anticipates that the
operator staffing plans will reflect a
simpler approach of showing that a
continuity of responsibility will be
maintained for facility operations
throughout the operating phase, with at
least one GLRO providing continuous
oversight and remaining immediately
available when any units are fueled.
Additionally, a revised approach to the
traditional position of the shift technical
advisor that focuses on the availability
of engineering expertise as a means of
addressing uncertainties and abnormal
circumstances is more suitable within
the context of part 53 and is intended
to be applicable to all facilities,
irrespective of other design and staffing
considerations.
Consistent with this approach, the
NRC proposes under § 53.730(f) to
require the submittal of a staffing plan
that details operations staffing, how
engineering expertise will be provided,
and what staffing will be available to
provide other needed support functions.
The NRC intends that this requirement
would be associated with staff guidance
for reviewing operations staffing plans
that is planned to accompany part 53
and that, following NRC approval of the
OL or COL, the staffing plan would
become a condition of the facility
license. The NRC intends that, at a
minimum, the approved licensed

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operator and senior operator (or, if
applicable, GLRO) staffing, positions,
and personnel locations will be
incorporated into corresponding
requirements within the facility TS and
that a license amendment would thus be
required for any subsequent changes.
Operator training and qualification
programs provide an essential
component of supporting human
performance in implementing tasks with
safety implications. Such programs
must include components that cover the
stages of initial training, examination,
and continuing training. Additionally,
recognizing the potential for varying
concepts of operations to affect
traditional, prescriptive approaches to
operator proficiency, the NRC proposes
under part 53 to allow facilities to
develop operator proficiency programs
based on facility-specific
considerations.
Therefore, the NRC proposes in
§ 53.730(g)(1) to require approval as part
of its approval of the OL or COL, of the
programs that will be used for the initial
training, initial examination,
requalification training and
examination, and proficiency of both
licensed operators and senior operators.
In a corresponding manner, the NRC
proposes in § 53.730(g)(2) to require
approval of the programs that will be
used for the GLRO equivalents of each
of these programs for facilities with
such staffing. The NRC intends that
examination program requirements
would be associated with staff guidance
for the review of tailored examination
processes that are planned to
accompany part 53. Following the
completion of an initial training
program, continuing training programs
provide an important means of
sustaining the knowledge and abilities
of individuals. The NRC is proposing to
adapt the requirements of § 50.54(i–1) in
§ 53.730(g)(3) to require that operator
continuing training programs be in
effect to support operator performance.
Under part 53, the NRC proposes to
require these programs to be in effect
concurrent with when the initial
operator examinations first commence,
in effect putting the programs in place
only when they are needed. This
represents a modification of the
comparable requirement of § 50.54(i–1),
which links the commencement of these
programs to a timeline driven by the
licensing of the facility.
The authorization to manipulate
controls of the facility that directly
affect reactivity or power level is
restricted to individuals who are either
licensed operators, licensed senior
operators, or GLROs. However, for
practical purposes, situations in which

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an individual is participating in an
approved training program or
reestablishing proficiency may also call
for them to operate the controls of the
facility under the cognizance of a
licensed individual. The NRC is
proposing to adapt the requirements of
§ 55.13 in § 53.735 to accomplish this,
with a notable difference being the
incorporation of GLROs.
Section 53.740 would provide
requirements for OL and COL holders
under part 53. Portions of § 53.740
would be adapted from the conditions
of § 50.54. In general, the conditions for
operations staffing under part 53 would
reflect considerations for potential
technological differences and varying
concepts of operation that are expected
among part 53 facility licensees.
Additionally, certain requirements
would be specific to the operating phase
while others would remain in effect
following the permanent cessation of
facility operations during the
decommissioning phase.
All commercial nuclear plants
licensed under part 53 would require
some form of licensed operator staffing,
whether it be by specifically or
generally licensed operators. Consistent
with this, the NRC is proposing under
§ 53.740(a) to require facility licensees
to demonstrate compliance with the
programmatic requirements for either
specifically licensed operators and
senior operators or for GLROs, as
applicable to the facility.
The NRC recognizes that technologyinclusive facility staffing will need to
account for a potentially wide range of
concepts of operations; for this reason,
flexible and performance-based
approaches for establishing required
facility staffing are appropriate.
However, once the appropriate facility
staffing has been determined and
approved by the NRC, such staffing
must be maintained to ensure that the
appropriately qualified individuals will
be available when needed to support the
safe operation of the facility. Therefore,
the NRC is proposing under § 53.740(b)
to require that the staffing described
within the approved facility staffing
plan be maintained as a condition of the
facility license as opposed to
prescriptive staffing requirements like
those of § 50.54(k) and (m).
Because operation of facility controls
directly affects reactivity or power level,
only those individuals who possess
appropriate levels of qualification and
authorization are permitted to operate
those controls. The NRC is proposing to
adapt the requirements of § 50.54(i) in
§ 53.740(c) to require that only
specifically licensed operators and
senior operators or, alternatively,

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GLROs, may operate facility controls,
with allowance for specified exceptions
for the purposes of operator training or
proficiency.
Senior operators, by virtue of their
license level, are qualified and
authorized both to perform certain
important responsibilities and to direct
the licensed activities of licensed
operators. Therefore, facilities that are
required to be staffed by specifically
licensed operators must also include
senior operators within their staffing. In
contrast, facilities staffed with GLROs
only have a single license level available
and, therefore, there is no equivalent
provision for such facilities. The NRC is
proposing to adapt the requirements of
§ 50.54(l) in § 53.740(d) to require the
licensing and designation of senior
operators at facilities staffed by
specifically licensed operators.
In contrast with control
manipulations that directly affect
reactor power and reactivity (e.g.,
control rod movement, control drum
rotation, recirculation pump speed
adjustment, reactor coolant system
boration or dilution, etc.) and are
therefore restricted to performance only
by licensed operators, other types of
plant operations that may result in
reactor power and reactivity changes via
means that are indirect in nature (e.g.,
electrical generation changes, turbine
bypass valve operation, steam usage by
process heat applications, etc.) may be
implemented by non-licensed
personnel. However, due to the
potential influence of such operations
on reactor power and reactivity, the
continuous oversight of reactor
parameters by a licensed operator is
necessary during these operations. The
NRC is therefore proposing to adapt the
requirements of § 50.54(j) in § 53.740(e)
to require appropriate oversight of
operations, other than those associated
with the controls themselves, that may
affect reactivity or power level.
Load following where plant output
automatically changes in response to
externally originated instructions or
signals is not permitted under the
existing regulations of § 50.54. However,
new technological considerations and
concepts of operation may justify such
an operational approach under
appropriate circumstances. The NRC
recognizes that, beyond electrical power
generation, load following may also
affect other applications of plant output,
such as hydrogen production,
desalination, or district heating. For
load following to be permissible,
measures must be in place to provide
assurance that plant output
considerations are not permitted to lead
to challenges to safe reactor operations.

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These measures may consist of
automated control systems, automatic
protective features, or the continuous
oversight and immediate intervention
capability of an appropriately qualified
and authorized individual. Section
53.740(f) would allow for load
following, provided that appropriate
measures are in place. In considering
the acceptability of the measures
associated with load following, the NRC
expects that any automatic protection
relied upon would be separate from that
credited for reactor protection purposes
and would employ setpoints that are set
so as to prevent actuation of the reactor
protection system while accomplishing
its functions to the extent practical.
Core alterations such as refueling are
associated with specific considerations
that warrant limiting the oversight of
such operations to appropriately
qualified and authorized individuals.
Unlike other types of fuel handling
operations, core alterations occur within
the confines of a reactor vessel that is
specifically designed to support and
sustain nuclear criticality, thereby
justifying the imposition of higher
qualification levels within such
contexts. The NRC is proposing to adapt
the requirements of § 50.54(m)(2)(iv) in
§ 53.740(g) to require the supervision of
core alterations by either a specifically
licensed senior operator, a specifically
licensed senior operator whose license
is limited to fuel handling, or by a
GLRO, as applicable to the facility.
Because certain commercial reactor
designs may be capable of refueling
while at power and, in any event,
overall facility oversight would already
be required by either a specifically
licensed senior operator or by a GLRO,
the NRC proposes to omit this
requirement as redundant during
periods where core alterations occur
while the plant is operating.
It is impossible to predict every
possible scenario that a commercial
nuclear plant might potentially
encounter. Therefore, it is prudent to
grant the authority for appropriately
qualified individuals to depart from
facility license conditions when
emergency circumstances dictate that
doing so is in the interest of public
health and safety. The NRC is proposing
to adapt the requirements of § 50.54(x)
and (y) in § 53.740(h) to permit specific
individuals to authorize departures from
facility license conditions or TSs when
emergency conditions warrant doing so
for the protection of the public health
and safety. Recognizing that certain
facilities licensed under part 53 may be
staffed by GLROs in lieu of specifically
licensed senior operators, the NRC
proposes to extend this authority to

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GLROs. While it is not anticipated that
GLROs will have a role in the
fulfillment of safety functions at selfreliant-mitigation facilities and,
furthermore, that operators at such
facilities would not be in a position by
which to significantly influence
radiological safety outcomes, the very
nature of the § 50.54(x) and (y) and the
proposed § 53.740(h) provisions concern
situations that are unanticipated and,
therefore, unforeseeable. Thus, it is
appropriate to grant GLROs a
comparable authority to that of senior
licensed operators and certified fuel
handlers as it relates to invoking this
provision under emergency conditions
as a means of accounting for such
possibilities.
Due to the unique authorities and
responsibilities of both specifically and
generally licensed reactor operators, it is
essential that any individual fulfilling
such a role demonstrate compliance
with the regulatory requirements for
operator licensing. Section 107 of the
Act authorizes the Commission to
prescribe conditions for the licensing of
operators and to issue licenses
consistent with those conditions. The
NRC is proposing to adapt the
requirements of § 55.3 in § 53.745 to
require that any person performing the
function of an operator, senior operator,
or GLRO must be authorized by a
license issued by the Commission.
The NRC proposes to license
individuals as operators under both
specific and general licensing
frameworks. Specific licenses would be
for licensed operators (i.e., reactor
operators) and senior operators (i.e.,
senior reactor operators) and would be
issued to a named person upon approval
by the Commission of an application for
that named person. In contrast, GLROs
would perform duties under the
provisions of a general license that
would be effective without the filing of
an application with the Commission or
the issuance of licensing documents to
a particular person. The NRC proposes
requirements for the use of a specific
licensing process for licensed operators
and senior operators under §§ 53.760
through 53.795, with § 53.760
addressing applicability.
Medical fitness is an important
component of the overall process of
specifically licensing operators because
it provides assurance that operators will
be able to carry out important duties
without being precluded from doing so
by health-related issues. Medical fitness
also provides assurance that such issues
will not adversely affect the
performance of assigned job duties or
cause operational errors that endanger
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a requirement for medical fitness, a
medical examination by a physician to
confirm compliance with this
requirement is necessary. The NRC is
proposing to adapt the requirements of
§§ 55.21, 55.23, and 55.27 under
§ 53.765 to require medical fitness,
examinations by physicians, and
medical certification for specifically
licensed operators and senior operators.
In recognition of the fact that GLROs are
not expected to have a role in the
fulfillment of safety functions at the
facilities at which they are licensed, the
NRC proposes to not extend a
comparable medical requirement to
GLROs.
The NRC is also proposing to adapt
the requirements of §§ 55.25 and
50.74(c) in § 53.770 to require that
timely notifications be made to the NRC
if a specifically licensed operator or
senior operator develops a permanent
physical or mental condition that
adversely affects the performance of
assigned operator job duties or could
cause operational errors endangering
public health and safety.
Notwithstanding this requirement
related to permanent medical
conditions, the NRC continues to
recognize that it is appropriate for
facility licenses to impose
administrative restrictions and
conditions upon specifically licensed
operators and senior operators in
response to temporary medical
conditions.
The process of specifically licensing
individuals as licensed operators or
senior operators requires the submittal
of applications to the NRC for review.
These applications must detail certain
elements associated with licensing,
including the demonstration of
compliance with examination,
experience, and medical requirements.
The NRC is proposing to adapt the
requirements of §§ 55.31 through 55.35
in § 53.775 to include requirements for
the applications associated with the
specific licensing of licensed operators
and senior operators at commercial
nuclear plants licensed under part 53. In
contrast with the part 55 requirements,
the NRC proposes to provide additional
flexibility by locating certain details
associated with the preparation and
submittal of these applications within
guidance in lieu of placement within
this proposed rule itself.
The NRC proposes overall
programmatic requirements for
specifically licensed operator and senior
operator training, examination, and
proficiency in § 53.780. In general, the
proposed requirements are adapted from
those in part 55, with several additional
flexibilities being incorporated to better

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account for potential variations in
reactor technologies and concepts of
operations. The requirements proposed
in § 53.780 cover, in part, the initial
training, initial examination,
requalification training, requalification
examination, and proficiency of
specifically licensed operators and
senior operators.
The initial training process provides
individuals with the knowledge and
abilities needed to subsequently fulfill
assigned duties as licensed operators or
senior operators in a safe and reliable
manner. The use of a systems approach
to training (SAT) ensures that the
training program is based upon job
requirements in a manner that can be
adapted to account for differences in
plant technology, concepts of
operations, and operator roles in the
fulfillment of design-specific safety
functions. The NRC is proposing under
§ 53.780(a) to require facility licensees
to implement a SAT-based training
program for the initial training of
licensed operator and senior operator
applicants. The program must be
adequate to ensure that applicants will
be capable of performing the duties
necessary both to protect public health
and safety and to maintain plant safety
functions. The NRC further proposes
that such programs be subject to NRC
approval and subsequent change control
processes of an appropriate nature.
Examinations provide a means of
assessing that individuals have achieved
a degree of knowledge and ability that
is sufficient to carry out assigned duties
as licensed operators or senior operators
in a manner that is safe and reliable.
The NRC is proposing to adapt the
requirements of §§ 55.40, 55.41, 55.43,
and 55.45 in § 53.780(b) to require that
facilities establish and implement an
initial examination program. However, a
key difference from the comparable
requirements of part 55 would be that
facilities have the flexibility to propose,
subject to NRC approval, the
examination methods and criteria to be
used in assessing satisfactory applicant
performance. Such examination
programs (including those used within
the scope of requalification training)
would need to provide for acceptable
levels of both test validity and test
reliability in order to be considered
acceptable. The NRC intends that staff
guidance would be available to facilitate
the review of licensing examination
programs that are proposed by facility
licensees and that, following NRC
approval, initial examination programs
would be subject to an appropriate
change control process. Furthermore,
the NRC proposes that holders of
licenses to operate commercial nuclear

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plants under part 53 be provided the
alternative of administering their own
approved licensing examinations. The
NRC would continue to exercise
appropriate oversight of the program,
make operator licensing decisions based
upon the examination results, and
reserve the right to administer the
examinations in lieu of permitting the
facility to do so. However, irrespective
of the provided flexibilities in
examination format and structure, at a
minimum, topics from the following
general categories of knowledge and
abilities should be sampled in such
examinations:
• Reactor Theory, Thermodynamics,
and Chemical Interactions
• Plant Systems and Components
• Reactivity Management and
Manipulations
• Radiation Control and Safety
• Emergency, Abnormal, and Normal
Operations
• Administrative Requirements and
Conditions of the Facility License
Requalification training programs
provide for the continuing training and
examination of specifically licensed
operators and senior operators to ensure
that they maintain the knowledge and
abilities needed to support the safe and
reliable performance of job duties
following the completion of an initial
training and examination program. The
NRC is proposing to adapt the
requirements of § 55.59 in § 53.780(c) to
require that facilities implement both a
SAT-based requalification training
program and a biennial requalification
examination program. However, a
notable difference from the biennial
requalification examinations required
under part 55 would be that distinct
annual operating test and biennial
written examination components would
not be mandated, with the facility
licensee instead proposing the
examination methods and criteria to be
used in assessing satisfactory
performance. The NRC intends that
guidance would be available to facilitate
the review of the requalification
examination programs that are proposed
by facility licensees and that, following
NRC approval, requalification
examination programs would be subject
to an appropriate change control
process.
For examinations to provide for valid
assessments of the knowledge and
abilities of individuals, the
examinations must remain free from
compromises that could affect their
underlying integrity. The NRC is
proposing to adapt the requirements of
§ 55.49 in § 53.780(d) to require that
examinations and related activities

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remain free from any compromise that
might affect the integrity of the
examination process.
Simulators provide a valuable means
of training and evaluating plant
operators, and the NRC is specifically
authorized under the Nuclear Waste
Policy Act of 1982, as amended
(NWPA), section 306 (42 U.S.C. 10226)
to establish regulations for the use of
simulators within such context. The
NRC is proposing to adapt the
requirements of § 55.46 in § 53.780(e) to
address the use of simulation facilities
for training, examinations, and
applicant experience requirements, as
well as to address the maintenance of
simulator fidelity. However, the
proposed requirements of part 53 would
not mandate that full scope, plantreferenced simulators be used and
would allow the use of alternative
simulation facilities consisting of, for
example, partial scope simulators or the
plant itself, provided that all associated
requirements can be demonstrated to be
met using alternative approaches and
methods. Additionally, in allowing for
the possibility that an applicant or
licensee might demonstrate compliance
with training, examination, or
experience requirements using the plant
itself, the NRC is not allowing the
initiation of transients on the actual
plant. Consistent with this, aside from
controlled reactivity manipulations that
are conducted for the purposes of
demonstrating compliance with
experience requirements, actual plant
components may not be operated for
these purposes. Rather, the NRC
perspective is that the use of the plant
for training and examination purposes
should be restricted to techniques such
as walkthroughs, job performance
measures, simulated tasks, use of
augmented reality technology, and
similar approaches that provide training
and examination value while avoiding
the operation of actual plant
components.
There may be situations in which
applicants for operator or senior
operator licenses have previous training
and experience that justifies waiving
some, or all, of the initial examination
requirements. The NRC is proposing to
adapt the requirements of § 55.47 in
§ 53.780(f) to allow for consideration of
requests for waivers of examinations
requirements. In contrast with the part
55 requirements, the NRC proposes to
locate certain details associated with
such waiver requests within guidance
documentation in lieu of placement
within the rule itself.
For licensed operators and senior
operators to perform their assigned
duties safely and reliably, it is essential

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that they perform those duties
frequently enough so as to maintain a
sufficient degree of proficiency. The
NRC is proposing to adapt the
requirements of § 55.53(e) and (f) in
§ 53.780(g) to require that specifically
licensed operators and senior operators
maintain proficiency and, if proficiency
is not maintained, regain proficiency
prior to resuming licensed duties.
However, in recognition of the fact that
varying concepts of operations are
possible for advanced reactor facilities,
the NRC is proposing, in contrast with
the requirements of part 55, to allow
facility licensees to establish their own
programs for operator proficiency,
subject to NRC approval.
As the holders of specific licenses,
licensed operators and senior operators
must be subject to license conditions on
an individual basis to ensure that the
basis upon which the licenses were
issued remains valid. The NRC is
proposing to adapt the requirements of
§ 55.53 in § 53.785 to require
appropriate conditions of licenses for
specifically licensed operators and
senior operators. However, in contrast
with the requirements of § 55.53(e) and
(f), the NRC is proposing to allow
certain aspects of operator proficiency
to be addressed by an NRC-approved
facility proficiency program.
Licenses for specifically licensed
operators and senior operators are
issued by the NRC and must remain
subject to modification or revocation.
The NRC is proposing to adapt the
requirements of §§ 55.51 and 55.61 in
§ 53.790 to address the issuance,
modification, and revocation of licenses
issued to specifically licensed operators
and senior operators.
The licenses issued to specifically
licensed operators and senior operators
are valid for a period of six years, after
which they expire, unless otherwise
renewed. The NRC is proposing to adapt
the requirements of §§ 55.55 and 55.57
in § 53.795 to address the expiration and
renewal of licenses issued to
specifically licensed operators and
senior operators.
In developing this proposed rule, the
NRC has discussed with stakeholders
the considerations that might justify the
omission of the specifically licensed
operators and senior operators.
However, even for an inherently safe
reactor with autonomous operation
features, certain important
administrative functions (e.g.,
compliance with TS, operability
determinations, NRC notifications,
emergency declarations, risk
assessment, maintenance oversight, and
radiological release limit compliance)
would still need to be accomplished by

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appropriately qualified and authorized
individuals. Additionally, the NRC
recognized that manual manipulations
of facility reactivity controls must only
be performed by individuals who have
been appropriately licensed by the
Commission. The NRC therefore
proposes under § 53.800 to establish a
new class of facility (defined as a selfreliant-mitigation facility), according to
the criteria contained in § 53.800 for
part 53. These facilities would employ
GLROs rather than specifically licensed
operators and senior operators. The
GLRO regulations offer enhanced
flexibilities and targeted relaxations in a
manner that is commensurate with the
modified role of such operators to
ensure the safe operation of the
associated facilities. In contrast, those
facilities not meeting the criteria of
§ 53.800 would instead be considered
interaction-dependent-mitigation
facilities and would require staffing by
specifically licensed operators and
senior operators. The terminology used
to designate these facility types reflects
differences in how operators are
anticipated to need to interact with their
plant systems in mitigating events and
achieving safe outcomes; such systems
may either need operators to interact
with them in some manner (i.e., be
interaction-dependent) or may instead
be able to rely fully upon their own
capabilities independent of operator
interaction (i.e., be self-reliant).
Generally licensed reactor operators
would differ from specifically licensed
operators because the latter would be
directly and independently evaluated by
the NRC as part of their licensing
process. This direct and independent
evaluation remains appropriate when
operators may reasonably be expected to
exert a significant influence on public
health and safety outcomes. Therefore, a
key determinant as to whether generally
licensed reactor operators can be
utilized in facility staffing is the
assessment of the operator’s role in
maintaining and fulfilling safety
functions at the facility, such as through
the performance of credited actions for
the mitigation of plant events.
The criteria proposed in § 53.800
would designate self-reliant-mitigation
facilities. These criteria are derived from
the following set of considerations:
• no human action needed to satisfy
radiological consequence criteria;
• no human action needed to address
LBEs;
• safety functions not allocated to
human action;
• reliance upon robust and highly
reliable safety features; and
• adequate defense in depth achieved
without reliance on human action.

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It should be noted that those facilities
not meeting the criteria proposed in
§ 53.800 would instead be classified as
interaction-dependent-mitigation
facilities and would require staffing by
specifically licensed operators and
senior operators instead.
Generally licensed reactor operators
would perform duties under the
provisions of a general license that
would be effective without the filing of
an application with the Commission or
the issuance of licensing documents to
a particular person. The NRC proposes
requirements for the general licensing
process for GLROs under §§ 53.805
through 53.820. The requirements for
GLROs would parallel those for senior
operators in regard to their comparable
administrative responsibilities.
Nonetheless, the requirements for
GLROs would be relaxed and
incorporate greater flexibilities
compared to the requirements for
specifically licensed operators in a
manner that is consistent with the
GLRO’s role in safety at self-reliantmitigation facilities.
In order to use GLROs in lieu of
specifically licensed operators and
senior operators, a OL/COL applicant
would need to demonstrate that its
proposed facility is a self-reliantmitigation facility, i.e., that it will
comply with the following requirements
on an ongoing basis: maintaining GLRO
qualifications for the performance of
important functions and tasks;
incorporating relevant programmatic
controls into TS; administering the
related programs for training,
examination, and proficiency; and
ensuring that the relevant provisions of
parts 26 and 73 are met. Additionally,
to provide for an accurate accounting of
what individuals are licensed under the
general license, facility licensees would
be required to report the identities of all
generally licensed reactor operators to
the NRC on an annual basis.
Furthermore, a facility licensee must
ensure that the facility design and
performance continue to meet the
technological criteria to be classified as
a self-reliant-mitigation facility (i.e., the
criteria of § 53.800) on a continual basis
during the operating phase, as the
relaxations afforded to such facilities in
the areas of operator licensing, staffing,
and HFE would be predicated on this
assumption. The NRC therefore
proposes under § 53.805 to establish
requirements for facility licensees that
address issues such as these. Finally,
the failure of a self-reliant-mitigation
facility to subsequently meet the criteria
of § 53.800 after the issuance of an OL
or COL would constitute a reportable
event (i.e., an unanalyzed condition that

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significantly degrades plant safety)
under the provisions of § 53.1630.
The NRC proposes the general license
for GLROs under § 53.810. GLROs
would be licensed as a class of
individuals under the provision of
§ 53.810(a) and would be subject to the
conditions specified in § 53.810(b)
through (g). Portions of these conditions
are adapted from § 55.53 and from those
conditions currently included in the
licenses issued to specifically licensed
operators and senior operators. The NRC
would retain the ability to suspend or
prohibit individuals from operating
under the general license should such
action be warranted.
The NRC proposes overall
programmatic requirements for GLRO
training, examination, and proficiency
under § 53.815. In general, these
proposed requirements are adapted from
those of part 55 and parallel those also
proposed for specifically licensed senior
operators in § 53.780. These
requirements include increased
flexibilities and several targeted
relaxations that reflect the limited role
of GLROs in facility safety. The
requirements proposed under § 53.815
cover, in part, the initial training, initial
examination, continuing training,
requalification examination, and
proficiency of GLROs. Section 53.805
would require the facility licensee to
develop, implement, and maintain these
programs. Section 53.810, in turn,
would prescribe that the requirements
of § 53.805 would need to be met as a
requirement of the general license. The
implication of this structure is that the
facility licensee would need to
implement these programs for training,
examination, and proficiency, and
GLROs would need to participate in
these programs to demonstrate
compliance with the requirements of the
general license.
The initial training process provides
GLROs with the knowledge and abilities
needed to fulfill assigned duties as
GLROs. The use of a SAT serves to
ensure that the training program is
based upon job requirements in a
manner that can be adapted to account
for differences in plant technology and
concepts of operations. The NRC is
proposing under § 53.815(b) to require
facility licensees to implement a SATbased training program for the initial
training of GLROs that is adequate to
ensure that they have the necessary
knowledge, skills, and abilities to
perform their duties. The NRC further
proposes that such programs would be
subject to NRC approval, oversight, and
appropriate change control processes.
The training program must ensure that

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GLROs maintain the necessary
knowledge, skills, and abilities.
Examinations provide a means of
assessing that individuals have achieved
a degree of knowledge and ability that
will be sufficient to enable them to carry
out assigned duties as GLROs in a
manner that is both safe and reliable.
The NRC proposes to adapt the
requirements of §§ 55.40, 55.41, 55.43,
and 55.45 in § 53.815(b) to require that
facility licensees establish and
implement an initial examination
program. A key difference from the
comparable requirements of part 55
would be that facility licensees would
be afforded the flexibility to propose,
subject to NRC approval, the
examination methods and criteria to be
used in assessing satisfactory individual
performance. Such examination
programs (including those used within
the scope of continuing training) would
need to provide for acceptable levels of
both test validity and test reliability in
order to be considered acceptable. The
NRC intends that staff guidance would
be available to facilitate the review of
initial examination programs that are
proposed by facility licensees and that
approved initial examination programs
would be subject to an appropriate
change control process. In contrast with
both the requirements of part 55 and the
proposed requirements of § 53.780, the
NRC does not intend to administer or
evaluate these initial examinations.
However, the examination processes
themselves will continue to be subject
to ongoing NRC oversight. Irrespective
of the provided flexibilities in
examination format and structure,
topics from the following general
categories of knowledge and abilities
should be sampled in such
examinations:
• Reactor Theory, Thermodynamics,
and Chemical Interactions
• Plant Systems and Components
• Reactivity Management and
Manipulations
• Radiation Control and Safety
• Emergency, Abnormal, and Normal
Operations
• Administrative Requirements and
Conditions of the Facility License
Continuing training programs provide
the ongoing training and examination of
GLROs to ensure that they maintain the
knowledge and abilities needed to
support the safe and reliable
performance of job duties following the
completion of an initial training and
examination program. The NRC is
proposing to adapt the requirements of
§ 55.59 in § 53.815(b) to require that
facility licensees implement both a
SAT-based continuing training program

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and a requalification examination
program. However, a notable difference
from the examinations required under
part 55 would be that distinct annual
operating test and biennial written
examination components would not be
mandated. The facility licensee would
instead propose examination methods
and criteria to be used in assessing
satisfactory performance. Furthermore,
unlike the comparable requirements of
part 55 and those proposed for
specifically licensed operators and
senior operators, a biennial periodicity
for requalification examinations would
not be prescribed. However, adequate
justification for the proposed periodicity
of requalification examinations would
be required. The NRC intends that staff
guidance would be available to facilitate
the review of the requalification
examination programs that are proposed
by facility licensees. Approved
requalification examination programs
would be subject to an appropriate
change control process.
For examinations to provide for valid
assessments of the knowledge and
abilities of individuals, the
examinations must remain free from
compromises that could affect their
underlying integrity. The NRC is
proposing to adapt the requirements of
§ 55.49 in § 53.815(d) to require that
examinations and related activities
remain free from any compromise that
might affect the integrity of the
examination process.
Simulators provide a valuable means
of training and evaluating plant
operators and the NRC is specifically
authorized under the NWPA, section
306 (42 U.S.C. 10226) to establish
regulations for the use of simulators
within such context. The NRC is
proposing to adapt the requirements of
§ 55.46 in § 53.815(e) to address the use
of simulation facilities for training and
examinations, and experience
requirements, as well as to address the
maintenance of simulator fidelity. The
use of full scope, plant-referenced
simulators would not be mandated. The
potential use of alternative simulation
facilities consisting of, for example,
partial scope simulators or the plant
itself, would be allowed provided that
all associated requirements could be
demonstrated to be met using
alternative approaches and methods.
Additionally, in allowing for the
possibility that an applicant or licensee
might demonstrate compliance with
training and examination requirements
using the plant itself, the NRC is not
allowing the initiation of transients on
the actual plant. Consistent with this,
aside from controlled reactivity
manipulations that are conducted for

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the purposes of demonstrating
compliance with experience
requirements, actual plant components
may not be operated for these purposes.
Rather, the use of the plant for training
and examination purposes should be
restricted to techniques such as
walkthroughs, job performance
measures, simulated tasks, use of
augmented reality technology, and
similar approaches that provide training
and examination value while avoiding
the operation of actual plant
components.
There may be situations in which
GLROs have previous training and
experience that justifies waiving some,
or all, of the initial examination.
Therefore, the NRC is proposing under
§ 53.815(f) to allow facility licensees to
waive some, or all, portions of initial
examinations provided that such
waivers are consistent with a program
that has been approved by the NRC.
For GLROs to safely and reliably
perform their assigned duties, it is
essential that they perform those duties
frequently enough so as to maintain a
sufficient degree of proficiency.
However, the NRC recognizes that
facilities that utilize GLROs may have
concepts of operation that warrant
unique proficiency considerations.
Therefore, the NRC is proposing in
§ 53.815(g) to require that facility
licensees develop, implement, and
maintain programs to maintain and
reestablish, if needed, the proficiency of
GLROs. This could occur, for example,
if an individual’s extended absence
from watch standing has rendered
proficiency requirements unmet.
The general license should remain in
effect for an individual only while that
individual remains employed in a
position that may call for the individual
to manipulate the reactivity controls of
the facility. The NRC proposes under
§ 53.820 to require that the general
license would cease to be applicable on
an individual basis when an
individual’s employment status
becomes such that this is no longer the
case. However, the NRC recognizes that
for some types of self-reliant-mitigation
facilities, very long periods may elapse
between circumstances that necessitate
manual manipulation of reactivity
controls. Therefore, the general license
remains in effect for an individual as
long as the individual’s current position
could potentially require that individual
to manipulate reactivity controls at
some point within the course of the
individual’s assigned job duties.
The NWPA, section 306 (42 U.S.C.
10226) authorizes and directs the NRC
to, in part, issue regulations and
guidance that address the training and

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qualifications of civilian nuclear power
plant operators, supervisors,
technicians, and other appropriate
operating personnel. The NRC
implements this in part 50 through the
requirements of § 50.120, ‘‘Training and
qualification of nuclear power plant
personnel.’’ The NRC is proposing
under § 53.830 to adapt, with
modifications, the requirements of
§ 50.120 for use in part 53 to provide
more flexible personnel training and
qualification requirements than those in
§ 50.120 and better reflect diverse
concepts of operations.
The NRC recognizes that the
categories of nuclear power plant
personnel in § 50.120 may not be
needed for the diverse concepts of
operations, staffing models, and nontraditional personnel roles and
responsibilities anticipated under
proposed part 53; conversely, and for
the same reasons, additional categories
of plant personnel may need to be
covered by part 53. The NRC also
recognizes that the timeframe prescribed
in § 50.120 for the establishment of
training programs may not be aligned
with the schedules associated with the
startup of certain types of commercial
nuclear plant facilities. However, the
NRC also recognizes that the SAT-based
training required under § 50.120
remains an appropriate means by which
training programs should continue to be
developed and implemented. Therefore,
the approach taken by the NRC in
addressing the training of certain plant
staff under the proposed part 53 reflects
greater flexibilities in personnel
categories and programmatic
timeframes, while still retaining the
requirement that such training programs
be based on SAT.
The NRC is proposing under § 53.830
to require SAT-based training programs
with the timeframe for when such
programs are required being based upon
when the associated personnel are
needed to support facility-specific
needs. The training programs would
cover the training and qualification of
plant personnel in the general categories
of supervisors, technicians, and other
appropriate operating personnel. The
licensee would not be required to seek
NRC approval of a training program
prior to usage. However, the licensee is
required to accommodate NRC
inspection of the training program. The
NRC intends to develop guidance to
facilitate the inspection of these training
programs but does not intend for such
guidance to preclude the potential for
the training programs to be maintained
by a separate, NRC-approved
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The proposed § 53.845 would require
programs to be developed,
implemented, and maintained to help
ensure that design features and human
actions have the capabilities and
reliabilities necessary to demonstrate
compliance with the safety criteria in
subpart B throughout the operating life
of each commercial nuclear plant. The
proposed programmatic requirements in
subpart F would also address areas such
as radiation protection needed to
control routine effluents during normal
operations. The proposed §§ 53.850
through 53.910 would require programs
to support specific activities needed to
ensure the prevention or mitigation of
unplanned events or to support normal
operations for any reactor design.
However, each holder of an OL or COL
would be required to assess whether
additional programs are needed for the
specific reactor design and location of
the commercial nuclear plant. Licensees
would be able to combine, separate, and
otherwise organize programs and related
documents as appropriate for the
technologies and organizations
associated with the commercial nuclear
plant.
Proposed § 53.850 would require a
radiation protection program associated
with the requirements in subparts B and
C for public doses resulting from normal
operations and the protection of plant
workers. The proposed requirements
related to doses from normal operations,
including routine effluents, would be
similar to those specified in § 50.36a,
‘‘Technical specifications on effluents
from nuclear power reactors,’’ and
related requirements in standard TS for
offsite dose calculation manuals. While
the proposed section would include
requirements that are technically and
programmatically similar to part 50,
proposed § 53.850 would not include a
requirement for effluent-related TS as is
required in § 50.36a. A proposed
requirement similar to that found in the
administrative controls section of TS for
operating reactors licensed under parts
50 and 52 would be included for
programmatic controls of solid wastes to
complement the design requirements in
proposed § 53.425.
Proposed § 53.855 would require an
emergency response plan that
demonstrates compliance with the
requirements in appendix E to part 50
and § 50.47(b) or § 50.160. The
regulations in § 50.47 stating that the
NRC will not issue certain licenses
unless it finds that there is reasonable
assurance that adequate protective
measures can and will be taken to
protect public health and safety in the
event of a radiological emergency apply
equally to applications under part 53

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86941

complying with the applicable
standards set forth in either § 50.160 or
the requirements in appendix E to part
50 and § 50.47(b).
In its 2008 Advanced Reactor Policy
Statement, the Commission stated their
expectation that ‘‘the safety features of
advanced reactor designs will be
complemented by the operational
program for Emergency Planning (EP).
This EP operational program, in turn,
must be demonstrated by inspections,
tests, analyses, and acceptance criteria
to ensure effective implementation of
established measures.’’ Consistent with
this policy statement, emergency plans
and emergency planning zones are not
safety features in the design. In SECY–
97–020, ‘‘Results of Evaluation of
Emergency Planning for Evolutionary
and Advanced Reactors,’’ dated January
27, 1997, the staff indicated that the
rationale upon which EP for current
reactor designs is based, that is,
potential consequences from a spectrum
of accidents, is appropriate for use as
the basis for EP for evolutionary and
passive advanced LWR designs and is
consistent with the Commission’s
defense-in-depth safety philosophy.
Also, in its Safety Goals Policy
Statement the Commission stated that:
‘‘A defense-in-depth approach has been
mandated in order to prevent accidents
from happening and to mitigate their
consequences. Siting in less populated
areas is emphasized. Furthermore,
emergency response capabilities are
mandated to provide additional defensein-depth protection to the surrounding
population.’’ Consistent with this policy
statement, proposed § 53.855
contributes an additional independent
layer of defense in depth for commercial
nuclear plants. Therefore, the
emergency plans and emergency
planning zones under proposed § 53.855
are not used to demonstrate compliance
with subpart B and subpart C of this
part. Rather, compliance with the
requirements in proposed § 53.855
would provide reasonable assurance
that adequate protective measures can
and will be taken to protect public
health and safety in the event of a
radiological emergency.
Proposed § 53.860 would identify the
applicable regulations for part 53
applicants related to the programs for
physical security, cybersecurity, FFD,
AA, and information security. These
programs are discussed in more detail in
section V, ‘‘Changes to Other Parts of 10
CFR,’’ of this document.
Proposed § 53.860(a) would establish
the physical protection program and
present a graded approach to physical
protection requirements. If a licensee
can meet the proposed criterion in

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§ 53.860(a)(2)(i), then the requirement to
protect against the design-basis threat
(DBT) of radiological sabotage would
not be applicable. The criterion in
§ 53.860(a)(2)(i) would require a
licensee to show that potential
consequences resulting from a DBT
initiated event would result in offsite
doses below the values in § 53.210 even
if licensee mitigation and recovery
actions, including any operator action,
are unavailable or ineffective. Where the
criterion is met, the resulting physical
protection requirements would be those
for protection of SNM and Category 1
and Category 2 radioactive material, if
applicable. This proposal would apply a
new regulatory approach for certain
commercial nuclear plants in which the
DBT of radiological sabotage would not
be applicable.
For those licensees able to meet the
criterion in § 53.860(a)(2), the NRC
would not conduct Force-On-Force
(FOF) exercise inspections. Section
170D.a of the Act permits the
Commission to determine which
licensed facilities are part of a class of
licensed facilities where NRCconducted FOF exercises are
appropriate to assess the ability of a
private security force of a licensed
facility to defend against any applicable
DBT. For the class of licensees that meet
the criterion of § 53.860(a)(2), it would
not be appropriate to conduct FOF
exercises to evaluate performance at
commercial nuclear plants where the
DBT of radiological sabotage is not
applicable and the facility poses a lower
risk to public health and safety from
potential radiation exposure. These
facilities would still have tailored
security requirements and oversight
consistent with their relatively low risk.
For those licensees not able to meet
the criterion in § 53.860(a)(2), proposed
§ 53.860(a) would permit the licensee to
choose one of two paths to provide
physical protection: (1) the current set
of requirements in § 73.55, which would
include any changes resulting from the
ongoing proposed rulemaking on
Alternative Physical Security
Requirements for Advanced Reactors 2
that provides pre-determined physical
security alternatives; or (2) the
performance-based requirements in
proposed § 73.100. In either case, the
licensee would be subject to NRCconducted FOF inspections.
Proposed § 53.860(b) would require
licensees to establish, implement, and
maintain an FFD program under part 26.
Section 53.860(c) would require
2 SECY–22–0072, ‘‘Proposed Rule: Alternative
Physical Security Requirements for Advanced
Reactors,’’ dated August 2, 2022.

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licensees to establish, implement, and
maintain an AA program in accordance
with either § 73.56 or proposed § 73.120,
as appropriate. Section 53.860(d) would
require licensees to establish,
implement, and maintain a
cybersecurity program in accordance
with either § 73.54 or proposed § 73.110.
Section 53.860(e) would require
licensees to establish, implement, and
maintain an information protection
system that complies with the
requirements of §§ 73.21, 73.22, and
73.23, as applicable.
Proposed § 53.865 would establish
requirements for quality assurance and
refer to appendix B of part 50 for the
part 53 requirements for SR design
features. Proposed requirements related
to evaluating and reporting changes to
the quality assurance program would be
included in proposed subpart I and
would be equivalent to those found in
§ 50.54.
The proposed § 53.870 would require
licensees to actively assess possible
degradation of SSCs from the effects of
aging, fatigue, and environmental
conditions. The proposed inclusion of
requirements related to designing and
monitoring for possible degradation
mechanisms reflects important lessons
learned from the history of LWRs and
the likely introduction of new design
features and materials in future
commercial nuclear plants. The
allowable combinations of design
features, operating experience, testing,
and monitoring during operations
would support performance-based
approaches to the initial licensing of
new technologies. The proposed
performance-based approach to integrity
assessment programs would also allow
for the subsequent consideration of
operating experience and appropriate
corrective actions or allowable
relaxations for ensuring that design
features comply with the proposed
functional design criteria of §§ 53.410
and 53.420. The proposed program
would be based upon a comprehensive
and integrated evaluation of the aging
and other degradation mechanisms
applicable to the design; identification
of the affected SSCs; the allowances
provided in the design of the SSCs for
degradation; and schedules and
procedures for determining if and at
what rate degradation is occurring, as
well as its cause. Risk insights could be
used to prioritize the monitoring,
evaluation, and management of
degradation based upon the importance
of the SSC to safety and the time frame
for when the effects of degradation
could be of concern.
Proposed § 53.875 would establish
requirements for a fire protection

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program supporting operations similar
to § 50.48. The proposed fire protection
program during operations would work
in concert with specific fire protection
requirements proposed in subpart C for
design and analyses and in proposed
subpart E for construction and
manufacturing.
Proposed § 53.880 would establish
requirements for an inservice inspection
(ISI) and inservice testing (IST) program,
which are historically important
activities conducted in accordance with
ASME codes and regulations in
§ 50.55a. While the proposed part 53
would not incorporate specific
consensus codes and standards into the
regulations, § 53.880 allows for the use
of generally accepted codes and
standards. The proposed requirement
for an ISI and IST program would
reinforce the need to develop
monitoring programs to be conducted
during a plant’s operations phase to
complement the design process and
address inherent uncertainties. The NRC
encourages the continued use of
consensus codes and standards
supporting design, testing, and
inspections to support integrated and
performance-based approaches in
demonstrating compliance with the
proposed requirements in part 53.
Proposed § 53.910 would establish
requirements for developing,
implementing, and maintaining
procedures (e.g., operations and
emergency operating procedures) and
guidelines (e.g., accident management
guidelines). The programmatic
requirements for many of the
procedures listed in this proposed
section would be similar to the
requirements found in the
administrative controls section of TS for
plants licensed under parts 50 and 52.
The proposed inclusion, where
appropriate, of accident management
guidelines in these requirements is
intended to ensure that an integrated set
of procedures and guidelines would be
established by licensees to ensure
command and control across the
spectrum of possible event sequences.
The proposed required procedures
would also include those needed to
complement the design requirements in
proposed § 53.440(m) related to
criticality alarms and the equivalent of
the procedures required in § 50.54(hh)
to address notifications of potential
aircraft threats.
Subpart G—Decommissioning
Requirements
The proposed subpart G would
provide the regulatory requirements for
the decommissioning phase of the life
cycle of a commercial nuclear plant.

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The requirements being proposed in
subpart G for the decommissioning of a
commercial nuclear plant are adapted
from the current regulations in § 50.75,
‘‘Reporting and recordkeeping for
decommissioning planning,’’ § 50.82,
‘‘Termination of license,’’ and § 50.83,
‘‘Release of part of a power reactor
facility or site for unrestricted use.’’
Although the requirements from those
sections of part 50 have been copied
into proposed subpart G with relatively
few changes, the requirements are
reorganized to fit within the part 53
structure. The few changes made were
primarily to make the proposed
requirements more technology inclusive
by adding alternatives within sections,
whereas some requirements in part 50
were developed specifically for LWRs.
As an example, § 50.75 provides
minimum amounts of decommissioning
funds required to demonstrate
reasonable assurance of funds for
decommissioning LWRs. Such generic
amounts have not been developed for all
reactor technologies that may be
licensed under part 53. Therefore, the
Commission proposes in § 53.1020,
‘‘Cost estimates for decommissioning,’’
that site-specific cost estimates for
decommissioning must be developed
considering costs in such areas as
engineering, labor, and waste disposal.
The derivation of the generic cost
estimates for LWRs in § 50.75 is
provided in NUREG/CR–5884, ‘‘Revised
Analyses of Decommissioning for the
Reference Pressurized Water Reactor
Power Station,’’ and NUREG/CR–6187,
‘‘Revised Analyses of Decommissioning
for the Reference Boiling Water Reactor
Power Station.’’ Similar to part 50, a
provision for an annual adjustment of
decommissioning cost estimates would
be included in proposed § 53.1030.
The NRC is currently pursuing
another rulemaking, ‘‘Regulatory
Improvements for Production and
Utilization Facilities Transitioning to
Decommissioning,’’ which was
published as a proposed rule for public
comment on March 3, 2022 (87 FR
12254). As these rulemakings progress,
the NRC will consider revisions to part
53 to align the two rulemaking efforts.
For example, the proposed § 53.1075
could be expanded to include or
reference requirements for
decommissioning in areas such as EP
and security in addition to the proposed
decommissioning fire protection plans
that would provide an equivalent to
§ 50.48(f).
Subpart H—Licenses, Certifications, and
Approvals
Proposed subpart H would provide
requirements related to applications

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under part 53 for NRC licenses,
certifications, or approvals for
commercial nuclear plants.
Proposed subpart H would specify
requirements applicable to all part 53
applications as well as requirements
specific to part 53 applications for
LWAs, ESPs, standard design approvals,
standard DCs, MLs, CPs, OLs, and COLs.
Proposed subpart H would be
equivalent to and include all existing
licensing, certification, and approval
processes currently covered under parts
50 and 52, with the exception of the
process for early review of site
suitability issues. Interactions with
external stakeholders during the
development of the proposed rule did
not identify significant interest in or
need for including the process for early
review of site suitability issues in part
53.
Much of the proposed subpart H
regulatory text is identical to the
corresponding language in parts 50 and
52, with minor changes to account for
cross references in part 53, to make
language technology neutral, or to
reflect the unique analytical approach in
part 53. In these instances, this
preamble discussion will describe the
language as ‘‘equivalent’’ to the existing
corresponding requirement in part 50 or
part 52 and will describe any
deviations, where applicable.
Because part 53 carries over the
majority of the licensing options from
parts 50 and 52, there are several
sections in proposed subpart H that are
similar to existing regulations in parts
50 and 52. Proposed § 53.1100 would
address filing of applications for
licenses, certifications, or approvals
under oath or affirmation and is
equivalent to § 50.30. The proposed
§ 53.1100 does not include the current
requirement in § 50.30(a)(2) that the
applicant maintain the capability to
generate additional copies, because it is
unnecessary in the age of electronic
submissions. In addition, the existing
requirement on applications for OLs in
§ 50.30(d) is included in proposed
§ 53.1124(g)(2), ‘‘Relationship between
sections,’’ covering OLs, rather than in
proposed § 53.1100.
Proposed § 53.1101 would lay out
activities requiring an NRC license and
is equivalent to § 50.10(b). Proposed
§ 53.1103 would address combining
applications and is equivalent to
§§ 50.31, 50.52, and 52.8. Proposed
§ 53.1103(b) would continue the
Commission’s practice of combining
multiple authorizations for a facility
under parts 30, 40, 50, 52, and 70 into
one license based on the Commission’s
authority under Section 161h. of the Act
to combine NRC licenses. Proposed

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§ 53.1106 would address elimination of
repetition and is equivalent to § 50.32.
Proposed § 53.1109 would provide
general information requirements for the
content of applications submitted to the
NRC under part 53 and is equivalent to
§ 50.33, with the exception of § 50.33(f)
on financial qualifications, which is
covered in proposed subpart J, and
§ 50.33(h) on earliest and latest dates for
completion of construction, which is
covered in § 53.1306 of this subpart.
Each application would need to include
information to address the items in
proposed § 53.1109 as cited in the
appropriate section of this subpart for
the application type.
One change from current
requirements can be found in proposed
§ 53.1109(i), which is not limited to
electricity generation as it is currently in
part 50. Some prospective NRC
applicants are considering development
of nuclear plants for other commercial
ventures, such as process heat
generation or hydrogen production. In
addition, § 53.1109(j), which requires
applications containing classified
information to separate that information
from the unclassified information in the
application, refers to ‘‘Restricted Data or
classified National Security
Information’’ instead of the term used in
the corresponding provision in
§ 50.33(j), ‘‘Restricted Data or other
defense information.’’ This change was
made to use the defined term in part 95
rather than ‘‘defense information’’ as
used in § 50.33(j). The usage in § 50.33(j)
dates back to the Atomic Energy
Commission amendment of that section
on January 19, 1956 (21 FR 355, 357)
and was not changed with the issuance
of part 95 (45 FR 14476; March 5, 1980)
after the establishment of the NRC and
the 1975 reissuance of the former
Atomic Energy Commission regulations.
The revised terminology also aligns
with its usage in § 53.1115.
Proposed § 53.1112 would address
environmental conditions and is
equivalent to § 50.36b. Proposed
§ 53.1115 would address requirements
for agreements limiting access to
classified information and is equivalent
to § 50.37. Proposed § 53.1118 would
address ineligibility of certain
applicants and is equivalent to § 50.38.
Proposed § 53.1120 would address
exceptions and exemptions from
licensing requirements for Department
of Defense and DOE facilities and is
equivalent to § 50.11. Proposed
§ 53.1121 would address public
inspection of applications and is
equivalent to § 50.39.
Proposed § 53.1124 would address the
relationship between the various
licenses, certifications, and approvals

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provided in this subpart, and the
requirements are equivalent to a number
of similar provisions in parts 50 and 52
including §§ 50.10, 52.13, 52.43, 52.73,
52.133, and 52.153. New provisions are
provided in § 53.1124(c) and (d), that
would allow an application for either a
standard design approval or a standard
DC under part 53 to reference applicable
licensing-basis information that
supported issuance of an OL or COL
under part 53. These provisions are
being proposed to offer additional
flexibility beyond what is currently
allowed under parts 50 or 52 for an
applicant who may wish to license a
first-of-a-kind reactor for operation prior
to seeking generic approval or
certification of the standard design.
Proposed § 53.1124(e) would address
the limitations that a manufactured
reactor may only be transported to a site
with a COL and is equivalent to
§ 52.153. Proposed § 53.1130 would
address LWAs and is equivalent to
§ 50.10.
Proposed §§ 53.1140 through 53.1188
would govern the content of ESP
applications. Proposed § 53.1140 is
equivalent to § 52.12. Proposed
§ 53.1143 would address filing of
applications and is equivalent to
§ 52.15. Proposed § 53.1144 would
address general information
requirements for the content of
applications and is equivalent to
§ 52.16.
Proposed § 53.1146 would specify
requirements for the technical contents
of applications and is equivalent to
§ 52.17. Proposed § 53.1146(b)(2)
provides applicants for ESPs a
regulatory option to propose major
features of the emergency plans or
complete and integrated emergency
plans in accordance with either the
requirements in § 50.160 of this chapter,
or the requirements in appendix E to
part 50 of this chapter and § 50.47(b) of
this chapter, as applicable.
Proposed § 53.1149 would address
standards for review of ESP applications
and administrative review of
applications, including hearings, and is
equivalent to §§ 52.18 and 52.21.
Proposed § 53.1155 would address
referral to the ACRS and is equivalent
to § 52.23. Proposed § 53.1158 would
address issuance of ESPs and is
equivalent to § 52.24. Proposed
§ 53.1161 would address the extent of
activities permitted and is equivalent to
§ 52.25. Proposed § 53.1164 would
address the duration of an ESP and is
equivalent to § 52.26. Proposed
§ 53.1167 would address provisions for
requesting a LWA after issuance of an
ESP and is equivalent to § 52.27.
Proposed § 53.1170 would address

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transfers of ESPs and is equivalent to
§ 52.28. Proposed § 53.1173 would
address applications for ESP renewals
and is equivalent to § 52.29. Proposed
§ 53.1176 would address criteria for
renewal of an ESP and is equivalent to
§ 52.31. Proposed § 53.1179 would
address the duration of an ESP renewal
and is equivalent to § 52.33. Proposed
§ 53.1182 would address the use of a
site for purposes other than those
described in the permit and is
equivalent to § 52.35. Proposed
§ 53.1188 would address finality of ESP
determinations and is equivalent to
§ 52.39.
Proposed §§ 53.1200 through 53.1221
would govern the contents of standard
design approval applications. Proposed
§ 53.1200 is equivalent to § 52.131.
Proposed § 53.1203 would address filing
of applications and is equivalent to
§ 52.135. Proposed § 53.1206 would
address general information
requirements for the content of
applications and is equivalent to
§ 52.136.
Proposed § 53.1209 would address
requirements for the technical content
of applications and is largely equivalent
to § 52.137. In proposed § 53.1209(a),
the NRC proposes text that expands the
discussion of ‘‘major portion’’ standard
design approvals. Additional discussion
regarding standard design approvals for
a major portion of a standard design can
be found in the NRC’s ‘‘A Regulatory
Review Roadmap for Non-Light Water
Reactors,’’ which considers the Nuclear
Innovation Alliance report ‘‘Clarifying
‘Major Portions’ of a Reactor Design in
Support of a Standard Design
Approval.’’ Proposed § 53.1209(b)
outlines the required content of the
Final Safety Analysis Report (FSAR).
Proposed requirements in
§ 53.1209(b)(2) for portions of the
application addressing design
information state that the application
must include design information
equivalent to that required for a
standard DC. This reference to the
pertinent DC requirements (specifically,
those in proposed § 53.1239(a)(2)
through (27)) is an efficiency that would
prevent the need to repeat many of the
same requirements for the content of a
standard design approval application.
Proposed § 53.1210 would address
requirements for the content of a
standard design approval application
other than the FSAR. Proposed
§ 53.1210(a) would require the inclusion
of a description of availability controls
that are not included in the FSAR.
Proposed § 53.1212 would address
standards for review of applications and
is equivalent to § 52.139. Proposed
§ 53.1215 would address referral to the

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ACRS and is equivalent to § 52.141.
Proposed § 53.1218 would address staff
approval of designs and duration of
design approvals and is equivalent to
§§ 52.143 and 52.147. Proposed
§ 53.1221 would address finality of
standard design approvals and
information requests and is equivalent
to § 52.145 with the exception that it
extends such finality to a standard
approval referenced in a DC application.
Standard design approvals issued to
date under part 52 have been issued
during the NRC’s review of the standard
DC application and have relied on the
same application content. However, a
future scenario could arise where the
DC application is not submitted until
after a design approval has been
granted. The NRC would apply the same
finality provisions in this situation as in
the situation where a standard design
approval is referenced in a COL
application.
There is no equivalent to proposed
§ 53.1221(d) in part 52 for standard
design approvals. This provision would
state that the Commission will require,
before granting a CP, COL, OL, or ML
which references a standard design
approval, that engineering documents
be completed and available for audit. A
similar provision is included in part 52
in relation to a standard DC; and the
NRC would require that design and
analysis information needed for the
Commission to make its safety
determination be complete and
available for any application the NRC is
reviewing. Making this explicit provides
increased clarity to future standard
design approval applicants under part
53.
Proposed §§ 53.1230 through 53.1263
would address standard DCs. Proposed
§ 53.1230 would address general
provisions for standard DCs and is
equivalent to § 52.41. Proposed
§ 53.1233 would address filing of
applications and is equivalent to
§ 52.45. Proposed § 53.1236 would
address general information
requirements for the content of
applications and is equivalent to
§ 52.46. Proposed § 53.1239 would
address requirements for the technical
content of applications and is
equivalent to § 52.47(a). The
requirements in proposed § 53.1239
have been modified from the analogous
requirements in § 52.47(a) to align with
the technical requirements in proposed
part 53.
Proposed § 53.1241 would address
requirements for the content of a
standard DC application other than the
FSAR and is equivalent to § 52.47(b)
and (d).

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Proposed § 53.1242 would address
review of applications and is equivalent
to §§ 52.48 and 52.51. Proposed
§ 53.1242(c) would include a provision
that would allow a DC applicant to
reference applicable licensing-basis
information for an OL or COL issued
under part 53. As explained previously,
this provision is being proposed to
explicitly allow flexibility for an
applicant who may wish to license a
first-of-a-kind reactor for operation prior
to seeking certification of the generic
reactor design. For NRC findings on a
reactor design in an OL or COL
proceeding, this proposal would
provide finality in a subsequent DC
application that references information
on the OL or COL proceeding’s docket.
This finality accorded to the OL or COL
findings would bind the NRC staff and
the ACRS but would not bind members
of the public or the Commission. (To the
extent an Atomic Safety and Licensing
Board (ASLB) might have a role in a DC
rulemaking, the OL or COL findings
would not bind the ASLB either.)
Specifically, members of the public
would have the opportunity to comment
on a proposed DC rule under wellestablished NRC practice. The rationale
for binding the NRC staff and ACRS is
similar to the rationale for a COL
applicant referencing a standard design
approval under part 52.
Proposed § 53.1245 would address
referral to the ACRS and is equivalent
to § 52.53. Proposed § 53.1248 would
address issuance of standard DCs and is
equivalent to § 52.54. Proposed
§ 53.1251 would address duration of
certifications and is equivalent to
§ 52.55(c). Proposed § 53.1254 would
address application for renewal and is
equivalent to § 52.57. Proposed
§ 53.1257 would address criteria for
renewal and is equivalent to § 52.59.
Proposed § 53.1260 would address
duration of renewals and is equivalent
to § 52.61. Proposed § 53.1263 would
address finality of standard DCs and is
equivalent to § 52.63.
Proposed §§ 53.1270 through 53.1291
would address MLs covering
manufacturing activities at one or more
licensee facilities. Proposed § 53.1270
would address the scope of these
sections and is equivalent to § 52.151.
Proposed § 53.1273 would address
filing of applications for an ML and is
equivalent to § 52.155(a).
Proposed § 53.1276 would address
general information requirements for the
content of ML applications and is
equivalent to § 52.156, with one
exception. Proposed § 53.1276 would
require each application for an ML to
also include the information required by
§ 53.1109(e). This information includes

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the type of license applied for, the use
to which the facility will be put, the
period of time for which the license is
sought, and a list of other licenses,
except operator’s licenses, issued or
applied for in connection with the
proposed facility to address the
potential variations in how MLs might
be formulated under the proposed part
53.
Proposed § 53.1279 would address
requirements for the technical content
of applications for MLs to be included
in the FSAR and is equivalent to
§ 52.157. In addition, the requirements
in proposed § 53.1279(a) and (b) have
been modified from the analogous
requirements in § 52.157 to align with
the technical requirements in proposed
part 53. Proposed § 53.1279(a)(2)
outlines the required content of the
application addressing design
information and states that the
application must include design
information equivalent to that required
for a standard DC. This reference to the
pertinent DC requirements is an
efficiency that would prevent the need
to repeat the same requirements for the
content of an ML application.
Proposed § 53.1279(c) would provide
application requirements related to the
deployment of the completed
manufactured reactor. Proposed
§ 53.1279(c)(1) would require inclusion
of information related to the procedures
governing the preparation of the
manufactured reactor for shipping to the
site where it is to be operated, the
conduct of shipping, and the
verification of the condition of the
shipped items upon receipt at the site.
Proposed § 53.1279(c)(2) would require
that the application include information
on the interaction of the design,
manufacture, and installation of a
manufactured reactor within the
applicant’s organization and the manner
by which the applicant will ensure close
integration between the designer,
contractors, and any licensee of a
facility in which the manufactured
reactor is to be installed. Finally,
proposed § 53.1279(c)(3) would require
that the application include a
description of the measures used for the
control of interfaces between the holder
of the ML and the holder of the COL for
the commercial nuclear plant at which
the manufactured reactor is to be
installed. This information is necessary
for the NRC to determine whether the
applicant would have appropriate
controls in place to ensure coordination
between parties involved in the design,
manufacture, and eventual operation of
any reactor manufactured under an ML.
Proposed § 53.1279(d) would include
additional requirements for application

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content for applicants seeking an ML for
manufactured reactors that will be
fueled at the factory under a 10 CFR part
70 license, consistent with the
requirements in § 53.620(d). These
provisions would require the
application to include information
related to loading fuel and the required
independent physical mechanisms to
prevent criticality and to otherwise
provide assurance that the fueled
manufactured reactor can be
successfully transported, installed, and
operated at a site for which the
Commission has issued a COL that
authorizes construction and operation of
a commercial nuclear plant using the
manufactured reactor.
Proposed § 53.1282 would provide
requirements for other application
content for MLs and is equivalent to
§ 52.158. Proposed § 53.1282(a)(1)
would provide requirements to include
in the ML application the ITAAC within
the scope of the ML that the COL holder
referencing the ML must satisfy.
Proposed § 53.1282(a)(2) would require
that the ITAAC from a referenced
standard design apply to the portions of
the ML design within the scope of the
referenced standard design. Proposed
§ 53.1282(a)(3) would state that the COL
application may include a notification
that required referenced standard DC
ITAAC have been satisfied at the
manufacturing facility.
Proposed § 53.1282(b) would require
an ML application to include an
environmental report and, consistent
with existing requirements, proposed
§ 53.1282(b)(2) would note that if the
ML application references a standard
DC, the environmental report need not
contain a discussion of severe accident
mitigation design alternatives for the
manufactured reactor as used in a
commercial nuclear plant.
Proposed § 53.1285 would provide
standards for review of applications and
administrative review of applications
for MLs, including hearings, and is
equivalent to §§ 52.159 and 52.163.
Proposed § 53.1286 would address
referral of applications to the ACRS and
is equivalent to § 52.165. Proposed
§ 53.1287 would address issuance of an
ML and is equivalent to § 52.167.
Proposed § 53.1288 would address
finality of MLs and is equivalent to
§ 52.171. Proposed § 53.1291 would
address the duration of MLs and is
equivalent to § 52.173. Proposed
§ 53.1293 would address the transfer of
MLs and is equivalent to § 52.175.
Proposed § 53.1295 would address the
renewal of MLs and is equivalent to
§§ 52.177, 52.179 and 52.181, with a
minor exception. Proposed
§ 53.1295(a)(3) would state that an ML

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for which a timely application for
renewal has been filed remains in effect
until the Commission has made a final
determination on the renewal
application, provided, however, that the
holder of an ML may not begin
manufacture of a manufactured reactor
less than six months before the
expiration of the license. The proposed
6-month time frame for this provision is
changed from the 3-year period in the
equivalent provision in part 52 because
future reactor applicants may present
smaller, simpler designs, to include
micro-reactor designs, in ML
applications than those that were
envisioned when the existing
requirements were written. A 6-month
time frame for this provision would
provide greater flexibility for ML
holders related to manufactured reactors
being produced when the ML expires.
Proposed §§ 53.1300 through 53.1348
would address licensing requirements
for CPs. Proposed § 53.1300 would set
out general requirements for CPs and is
equivalent to § 50.23. Proposed
§ 53.1306 would address the general
information requirements for the
content of applications for CPs and is
equivalent to § 50.33(f) and (h).
Proposed § 53.1309 would address
requirements for the technical content
of applications for CPs and includes the
requirement to submit a Preliminary
Safety Analysis Report (PSAR) that
describes the facility and presents a
preliminary safety analysis of the
facility as a whole. This is in contrast to
an OL application which is required to
include an FSAR that describes the
facility and presents a final safety
analysis of the facility as a whole.
Proposed § 53.1309 is equivalent to
§ 52.17(a)(1)(iv) through (a)(1)(x) and
52.17(b), with two exceptions. First,
proposed § 53.1309 would replace the
analysis of the dose criteria required by
§ 52.17(a)(1)(ix) with analysis to
demonstrate compliance with the safety
criteria defined in §§ 53.210 and 53.220.
Second, proposed § 53.1309(a)(2) would
add a requirement for a CP application
to include several categories of detailed
design information, although
§ 53.1309(a)(2)(ii) would allow certain
relaxations of this requirement in view
of aspects of a design that may not yet
be fully developed. Section 53.1309
would reference the requirements for
the content of an ESP application to
address application requirements
related to siting and would reference the
requirements for the content of a DC
application to address application
requirements related to design of the
commercial nuclear plant. Proposed
§ 53.1309(a)(2)(ii) would address the
treatment of preliminary design

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information and notes that information
provided in the application may include
some aspects of the design that are not
fully developed. This provision would
require that the completed design,
including any changes during
construction, be described in the FSAR
in an application for an OL. This would
include the requirement for a
description of the PRA required by
§ 53.450(a) and its results. Probabilistic
risk assessments developed for
commercial nuclear plants prior to
construction would be based on the
design and other information available
at the time of the CP application. PRAs
performed in early design stages or prior
to construction may be inherently less
detailed and may include projected
information that will be subsequently
verified or revised when the plant is
built. Proposed § 53.1309(a)(4) would
address preliminary description of the
plans for coping with emergencies.
Proposed § 53.1312 would address
other application content for CPs.
Proposed § 53.1312(a)(1) is equivalent to
§ 52.80(b) but is adapted for a CP
application. Proposed § 53.1312(a)(2) is
equivalent to § 52.80(c) but is adapted
for a CP application. Proposed
§ 53.1312(b)(1) is equivalent to
§ 52.79(b), (c), and (d) but is adapted for
a CP application. Section 53.1312(b)(2)
is equivalent to portions of
§§ 52.63(b)(1), 52.79(b)(1) through (b)(3),
(c), and (d)(1) and (d)(3), 52.80, and
52.93(b), but is adapted for a CP
application. Guidance for equivalent
requirements in parts 50 and 52 is also
addressed in RG 1.206, ‘‘Applications
for Nuclear Power Plants,’’ Revision 1,
section C.1.7.
Proposed § 53.1315 would address
standards for review of applications and
administrative review of applications,
including hearings, and is equivalent to
§§ 52.81 and 52.85, but is adapted for a
CP application.
Proposed § 53.1318 would address
finality of NRC approvals, licenses, and
certifications referenced in a CP
application and is equivalent to
§ 52.83(a) but is adapted for a CP
application.
Proposed § 53.1324 would address
referral to the ACRS and is equivalent
to § 50.58(a) and to § 52.87 but is
adapted for a CP application.
Proposed § 53.1327 would address
authorization to conduct LWA activities
and is equivalent to § 52.91 but is
adapted for a CP application. Proposed
§ 53.1327(a) is equivalent to § 52.91(a)
but is adapted for a CP application.
Proposed § 53.1327(b) is equivalent to
§ 52.91(b) but is adapted for a CP
application. Proposed § 53.1330 would

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address exemptions, departures, and
variances for CP applicants.
Proposed § 53.1333 would address
issuance of CPs. Proposed § 53.1333(a)
is equivalent to § 50.35(a). Proposed
§ 53.1333(b) is equivalent to § 50.35(b)
and to § 52.97(c) but is adapted for a CP
application. Proposed § 53.1336 would
address the effect of CPs and is
equivalent to § 50.35(b). Proposed
§ 53.1342 would address the duration of
CPs. Proposed § 53.1342(a) is equivalent
to § 50.55(a). Proposed § 53.1342(b) is
equivalent to § 50.55(b). Proposed
§ 53.1345 would address the transfer,
assignment, and disposal of CPs and is
equivalent to § 50.80. Proposed
§ 53.1348 would address the
termination of CPs and is equivalent to
§§ 52.3(b)(8) and 52.110(a)(1) but is
adapted for a CP application.
Proposed §§ 53.1360 through 53.1405
address requirements for OLs.
Proposed § 53.1366 would address
requirements for the general content of
applications for OLs. It would refer to
general content requirements in
proposed § 53.1109 and would require
supplemental information. Proposed
§ 53.1366(a) is equivalent to § 50.33(f).
Proposed § 53.1366(b) is equivalent to
§ 50.33(k).
Proposed § 53.1369 would provide
requirements for the technical content
of applications for OLs to be included
in the FSAR and is equivalent to
§ 50.34(b) but has been modified to align
with the technical requirements in part
53. It would require that the FSAR
include and, as needed, update
information provided in the PSAR that
was submitted and reviewed to support
the associated CP application.
Similar to the proposed requirements
for the content of CP applications,
proposed § 53.1369(a) would reference
the requirements for the content of an
ESP application to address application
requirements related to the site. Section
53.1369(b) would reference the
requirements for the content of a DC
application to address some of the
application requirements related to
design of the commercial nuclear plant.
Proposed § 53.1369(c) is equivalent to
§ 50.34(b)(7). Proposed § 53.1369(d)
would require a description of the
Integrity Assessment Program that
would be required by proposed
§ 53.870. Proposed § 53.1369(e) is
equivalent to § 50.34(e). Proposed
§ 53.1369(g) would provide
requirements for OL application content
to support proposed § 53.730 related to
the role of personnel in the operation of
the commercial nuclear plant and is
adapted from requirements in part 55
and § 50.34(f). Likewise, proposed
§ 53.1369(h) would provide

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requirements for OL application content
related to training programs to support
proposed §§ 53.730(g) and 53.830 and
includes requirements equivalent to
§ 50.34(b)(8), § 52.79(a)(33), and part 55.
Proposed § 53.1369(i) would provide
requirements for OL application content
related to emergency plans to support
proposed § 53.855 and is equivalent to
§ 50.34(b)(6)(v).
Proposed § 53.1369(j) would provide
requirements for OL application content
related to the applicant’s organizational
structure and is equivalent to
§ 50.34(b)(6)(i). Proposed § 53.1369(k)
would provide requirements for OL
application content related to the
applicant’s proposed maintenance
program to support proposed § 53.715
and is equivalent to § 50.34(b)(6)(iv).
Proposed § 53.1369(l) would provide
requirements for OL application content
related to the applicant’s quality
assurance program to support proposed
§ 53.865 and is equivalent to
§ 50.34(b)(6)(ii). Proposed § 53.1369(m)
would provide requirements for OL
application content related to the
applicant’s proposed radiation
protection program to support proposed
§ 53.850 and is equivalent to
§ 50.34(b)(3).
Proposed § 53.1369(n) through (p)
would provide requirements for OL
application content related to the
applicant’s proposed physical security
program to support proposed § 53.860(a)
and are equivalent to § 50.34(c) and (d).
Proposed § 53.1369(q) would provide
requirements for OL application content
related to the applicant’s proposed
cybersecurity plan to support proposed
§ 53.860(d) and is equivalent to
§§ 52.79(a)(36)(iv) and 73.54. Proposed
§ 53.1369(r) would provide
requirements for OL application content
related to the implementation of
proposed security, safeguards, and
cybersecurity plans to support proposed
§ 53.860 and is equivalent to
§ 52.79(a)(35)(ii) and 52.79(a)(36)(iv)
and (v).
Proposed § 53.1369(s) would provide
requirements for OL application content
related to the applicant’s proposed fire
protection program to support proposed
§ 53.875 and is equivalent to
§ 52.79(a)(40). Proposed § 53.1369(t)
would provide requirements for OL
application content related to the
applicant’s proposed ISI and IST
program to support proposed § 53.880
and is equivalent to part of
§ 52.79(a)(11). Proposed § 53.1369(w)
would provide requirements for OL
application content related to the
applicant’s general employee training
program to support proposed § 53.830
and is equivalent to § 52.79(a)(33).

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Proposed § 53.1369(x) would provide
requirements for OL application content
related to the applicant’s FFD program
to support part 26 and is equivalent to
§ 52.79(a)(44). Proposed § 53.1369(y)
would provide requirements for OL
applicant’s programs to demonstrate
that any safety questions identified at
the CP stage have been resolved and is
equivalent to § 50.34(b)(5). Proposed
§ 53.1369(z) would provide
requirements for OL applicants to
describe how the performance of each
safety design feature has been
demonstrated capable of fulfilling
functional design criteria considering
interdependent effects through either
analysis, appropriate test programs,
prototype testing, operating experience,
or a combination thereof to support
proposed § 53.440(a). It is largely
equivalent to §§ 50.34(b)(5) and
50.43(e). Proposed § 53.1369(aa) would
provide requirements for OL application
content related to the applicant’s
proposed TS to support proposed
§ 53.710(a) and is equivalent to
§ 50.34(b)(6)(vi).
Proposed § 53.1372 would address
requirements for the content of OL
applications other than the FSAR.
Proposed § 53.1372(a) would require
submission of an environmental report
and is equivalent to § 50.30(f) and
§ 51.53(b). Proposed § 53.1372(b) does
not have a direct parallel in parts 50 and
52 and would require the inclusion of
a description of availability controls that
are not included in the FSAR to support
proposed § 53.710(b).
Proposed § 53.1375 would address
standards for review of OL applications
and the administrative review of
applications, including hearings, and is
equivalent to §§ 52.81 and 52.85, except
that the NRC has omitted 10 CFR part
54, ‘‘Requirements for Renewal of
Operating Licenses for Nuclear Power
Plants,’’ from the list of standards in the
proposed § 53.1375(a). Proposed part 53
does not include detailed requirements
related to renewal of licenses, although
a general provision and possible
placeholder for future requirements has
been included as proposed § 53.1595.
The NRC will decide after the part 53
final rule is published whether this
future section will be retained in part 53
to address license renewal or whether
the agency will take another approach to
address license renewal for part 53
licensees, such as amending part 54 to
address part 53 licensees.
Proposed § 53.1381 would address
referral to the ACRS and is equivalent
to §§ 50.58 and 52.87. Proposed
§ 53.1384 would address exemptions,
departures, and variances for OL
applicants. Section 53.1384(a) is

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equivalent to § 52.93 but is adapted for
OLs. Proposed § 53.1384(b) is equivalent
to §§ 52.39(d) (with respect to ESPs) and
52.93 but is adapted for OLs.
Proposed § 53.1387 would address
issuance of OLs. The proposed
introductory paragraph is equivalent to
§ 50.56. Proposed § 53.1387(a)(1)(i) is
equivalent to §§ 50.50 and 50.57(a)(1).
Proposed § 53.1387(a)(1)(ii) is
equivalent to § 50.50. Proposed
§ 53.1387(a)(1)(iii) is equivalent to
§ 50.57(a)(2). Section 53.1387(a)(1)(iv) is
equivalent to § 50.57(a)(3). Proposed
§ 53.1387(a)(1)(v) is equivalent to
§ 50.57(a)(4). Proposed
§ 53.1387(a)(1)(vi) is equivalent to
§ 50.57(a)(6). Proposed
§ 53.1387(a)(1)(vii) is equivalent to
§ 50.57(a)(5). Proposed
§ 53.1387(a)(1)(viii) is equivalent to
§ 52.97(a)(1)(vi) but is adapted for OLs.
Proposed § 53.1387(c) is equivalent to
§ 50.57(b). Proposed § 53.1387(d) is
equivalent to §§ 50.36(b) and 50.50.
Proposed § 53.1390 would address
backfitting of OLs and is equivalent to
§ 52.98(a) but adapted for an OL
application. Proposed § 53.1396 would
address duration of an OL and is
equivalent to § 50.51(a) and § 52.104.
Proposed § 53.1399 would address
transfer, assignment, and other
disposition of an OL and is equivalent
to § 50.80. Proposed § 53.1402 would
address applications for renewal of an
OL and refers to proposed § 53.1595.
Proposed § 53.1405 would address
continuation of an OL and is equivalent
to § 52.109 but is adapted to address an
OL.
Proposed §§ 53.1410 through 53.1461
would address requirements for COLs.
Proposed § 53.1410 is equivalent to
§ 52.71. Proposed § 53.1413 would
address general information
requirements for the content of
applications for COLs and is equivalent
to § 52.77, which references § 50.33.
Most of the provisions from § 50.33 are
restated in proposed § 53.1109. Some
requirements in § 50.33 related to
financial qualifications and construction
timelines are addressed in other
sections of part 53.
Proposed § 53.1416 would address the
technical content to be included in an
FSAR for an application for a COL and
is equivalent to § 52.79 except as
modified to reflect the technical
requirements in part 53 and with one
addition. Proposed § 53.1416 includes
the statement that the Commission will
require, before issuance of a COL, that
engineering documents, such as
analyses, drawings, procurement
specifications, or construction and
installation specifications, be completed
and available for audit if the more

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detailed information is necessary for the
Commission to verify the information in
the application and make its safety
determination. This statement is
equivalent to DC application
requirements in § 52.47 and is included
in proposed § 53.1416 for clarity.
Similar to the proposed requirements
for the content of OL applications,
proposed § 53.1416(a)(1) would
reference the requirements for the
content of an ESP application to address
application requirements related to
siting. Section 53.1416(a)(2) would
reference the requirements for the
content of a DC application to address
some of the application requirements
related to design of the commercial
nuclear plant. The remaining items
under proposed § 53.1416(a) are
likewise similar to the required content
for OL applications under proposed
§ 53.1369(a). Proposed § 53.1416(b)
would require COL applicants to
provide a report documenting the
resolution of any safety questions for
SSCs for which research and
development was necessary to confirm
the adequacy of their design and is
equivalent to § 50.34(b)(5). Proposed
§ 53.1416(c) would provide
requirements for COL applicants to
describe how the performance of each
safety design feature has been
demonstrated capable of fulfilling
functional design criteria considering
interdependent effects through either
analysis, appropriate test programs,
prototype testing, operating experience,
or a combination thereof to support
proposed § 53.440(a). It is largely
equivalent to §§ 52.79(a)(24) and
50.43(e). Proposed § 53.1416(d) would
address the content of COL applications
referencing an ESP. Proposed
§ 53.1416(e) would address the content
of COL applications referencing a
standard design approval. Proposed
§ 53.1416(f) would address the content
of COL applications referencing a
standard DC. Proposed § 53.1416(g)
would address the content of COL
applications referencing an ML.
Proposed § 53.1419 would address
other application content for COLs and
is equivalent to § 52.80. Proposed
§ 53.1419(a)(2) is new and would
require the inclusion of a description of
availability controls that are not
required to be included in the FSAR.
Proposed § 53.1422 would address
standards for review of applications and
the administrative review of
applications, including hearings, and is
equivalent to §§ 52.81 and 52.85. The
NRC has removed part 54 from the list
of standards in proposed § 53.1422(a).
Proposed part 53 does not include
requirements related to renewal of

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licenses, in relation to proposed
§§ 53.1422 and 53.1595.
Proposed § 53.1425 would address the
finality of NRC approvals referenced in
a COL application and is equivalent to
§ 52.83(a). Proposed § 53.1431 would
address the referral of COL applications
to the ACRS for review and is
equivalent to § 52.87. Proposed
§ 53.1434 would address the
authorization to conduct LWA activities
and is equivalent to § 52.91. Proposed
§ 53.1437 would address exemptions,
departures, and variances and is
equivalent to § 52.93. Proposed
§ 53.1440 would address issuance of
COLs and is equivalent to § 52.97.
Proposed § 53.1443 would address
finality of COLs and is equivalent to
§ 52.98.
Proposed § 53.1449 would address
inspection during construction and is
equivalent to § 52.99. Proposed
§ 53.1452 would address operation
under a COL and is equivalent to
§ 52.103. Paragraph (a) of proposed
§ 53.1452 would include footnotes to
provide that, for licensees installing
fueled manufactured reactors under a
COL, (1) the COL holder would notify
the NRC of its scheduled date for
initiating the physical removal of any
one of the independent physical
mechanisms to prevent criticality
required under § 53.620(d)(1) rather
than its scheduled date for the initial
loading of fuel, and (2) the NRC would
time its publication of the notice of
intended operation based on the COL
holder’s schedule for initiating the
physical removal of any one of the
independent physical mechanisms to
prevent criticality required under
§ 53.620(d)(1) rather than the COL
holder’s scheduled date for the initial
loading of fuel. These footnotes are
consistent with the provisions of
proposed § 53.620(d)(1)(iv), which
would state that, upon initiating the
physical removal of any one of the
independent physical mechanisms to
prevent criticality in the manufactured
reactor’s place of operation, the fueled
manufactured reactor has commenced
operation. For reactors without the
independent physical mechanisms to
preclude criticality under proposed
§ 53.620(d)(1), operation begins with
initial fuel load. In both cases, removal
of the physical features to prevent
criticality (for reactors with such
features) and initial fuel load (for
reactors without such features) put a
fully constructed utilization facility in a
position to sustain a nuclear chain
reaction, and in both cases, the
utilization facility cannot sustain a
nuclear chain reaction (for lack of
sufficient reactivity) until the action

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takes place. Therefore, the NRC
proposes that initiating the physical
removal of any one of the independent
physical mechanisms to prevent
criticality is the best analogue to initial
loading of fuel for reactors without such
features.
The proposed footnote in § 53.1452(a)
regarding timing of the notice of
intended operation for fueled
manufactured reactors with
independent physical mechanisms to
prevent criticality also addresses the
requirements of Section 189a.(1)(B)(i) of
the Act. This section requires, in part,
that ‘‘[n]ot less than 180 days before the
date scheduled for initial loading of fuel
into a plant by a licensee that has been
issued a combined construction permit
and operating license under section
185b., the Commission shall publish in
the Federal Register notice of intended
operation.’’ That section further requires
that this notice provide a 60-day period
in which to request a hearing ‘‘on
whether the facility as constructed
complies, or on completion will
comply, with the acceptance criteria of
the license.’’ In the case where a fueled
manufactured reactor arrives at the site
where it is to be operated by a COL
holder, the manufacturer would have
loaded fuel at the factory under its part
70 license. Therefore, at the site of
operation, there would not be ‘‘initial
loading of fuel into a plant by a licensee
that has been issued a combined
construction permit and operating
license’’ (emphasis added). Under a
literal reading of the entry condition in
Act Section 189a.(1)(B)(i), this situation
would not trigger its requirements.
However, the purpose of the provision
is to offer the hearing opportunity at
least 180 days prior to when the fuel is
loaded and ready for use at its
authorized location. It would be
contrary to that purpose if, in this
situation, the Commission did not
publish the notice of intended operation
and opportunity for the public to
request a hearing on conformance with
the acceptance criteria in the COL for
the site of operation. To fulfill the
underlying purpose of the law, the NRC
proposes to time the notice of intended
operation based on the COL holder’s
schedule for initiating the physical
removal of any one of the independent
physical mechanisms to prevent
criticality required under § 53.620(d)(1).
This action by the COL holder would be
the best analogue to initial fuel load by
the COL holder for the reasons stated
previously. This analogue is adopted in
other sections of the proposed part 53
and related sections in parts 50 and 73
that use initial fuel loading to identify

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a transition point for the applicability of
regulatory requirements. To address the
possible loading of fuel into a
manufactured reactor for subsequent
transport to and use at a commercial
nuclear plant, multiple sections that
determine the applicability of
regulations have been drafted or revised
to allow for either initial fuel load or
initiating the physical removal of any
one of the independent physical
mechanisms to prevent criticality
required under § 53.620(d)(1) for a
fueled manufactured reactor to
determine the applicability of the
requirement, as appropriate.
Proposed § 53.1455 would address
duration of COL and is equivalent to
§ 52.104. Proposed § 53.1456 would
address the transfer of a COL and is
equivalent to § 52.105. Proposed
§ 53.1458 would address application for
renewal and is equivalent to § 52.107.
Proposed § 53.1461 would address
continuation of COL and is equivalent
to § 52.109.
Proposed § 53.1470 would address
standardization of commercial nuclear
plant designs and licenses to construct
and operate commercial power reactors
of identical design at multiple sites and
is equivalent to appendix N of part 52.
This section would set out the particular
requirements and provisions applicable
to situations in which applications for
CPs and subsequent OLs, or COLs,
under this part are filed by one or more
applicants for licenses to construct and
operate nuclear power reactors of
identical design (‘‘common design’’) to
be located at multiple sites. Additional
information related to this proposed
section is provided in the final rule to
revise part 52 (72 FR 49352; August 28,
2007).
Subpart I—Maintaining and Revising
Licensing-Basis Information
Part 53 would establish requirements
for the maintenance of licensing-basis
information in subpart I.
Section 53.1500 would describe the
purpose of the subpart in terms of the
definition of licensing-basis information
in subpart A. Subpart I would be closely
tied to the requirements in subpart H,
which would provide the requirements
for contents of applications for the
various types of licenses issued under
part 53. Subpart I would generally be
organized into sections dealing with: (1)
licensing-basis information that
licensees are not authorized to change
without NRC approval (e.g., licenses,
regulations); and (2) licensing-basis
documents that licensees may change
provided specified criteria are satisfied
(e.g., FSAR, program descriptions). The
subpart would also capture certain

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general conditions on licenses and
changes to the licenses related to the
transfer and termination of licenses.
Section 53.1502 would define specific
terms and conditions of licenses. These
terms and conditions would be
equivalent to the regulations in: (1)
§ 50.54(h) stating that each license is
subject to the provisions of the Act and
requirements issued by the Commission;
(2) § 50.54(s) stating the actions the
Commission would take if it makes a
finding that there is not reasonable
assurance that adequate protective
measures can and will be taken in the
event of a radiological emergency; (3)
§ 50.54(aa) stating that each license is
subject to the specified sections of the
Federal Water Pollution Control Act;
and (4) § 50.54(dd) stating that a holder
of an OL or COL may take reasonable
actions that depart from the license in
a national security emergency.
Section 53.1505(a) would serve as an
introduction to and overview of the
sections that follow on changes to
licensing-basis information requiring
prior NRC approval, namely the
elements of licensing-basis information
defined by licenses, orders, and
regulations. The related sections within
these subparts would primarily deal
with the process of how a licensee
requests and the NRC issues an
amendment to a license or issues an
order that modifies a license. Another
important element of licensing-basis
information that a part 53 licensee
would not be able to change or deviate
from without NRC approval would be
the NRC regulations themselves. Section
53.1505(b) would refer to § 53.080 in
subpart A that would provide the
criteria for a licensee or other party to
satisfy when requesting an exemption
from NRC regulations.
Section 53.1510 would be equivalent
to § 50.90 and would require that a
licensee submit an application to
request an amendment to a license. The
required assessments that would be
included within an application to
amend a license under part 53 would
need to address the safety criteria and
analysis requirements of subparts B and
C. As with parts 50 and 52, licensees
would be required to include in their
applications to amend a license an
analysis of whether the amendment
involves no significant hazards
consideration using the standards in
§ 53.1520, which would be equivalent to
the standards in § 50.92. Although this
rulemaking provided an opportunity to
revise the terminology related to no
significant hazards consideration
determinations, which dates to the early
1960s when applications were
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reports, the NRC is proposing to
maintain the same terminology used in
part 50 to minimize the need for
associated changes in other regulations,
guidance, and public notices.
Section 53.1515 would establish
requirements for public notices and
state consultations associated with the
NRC’s processing of a license
amendment request. This section would
be equivalent to § 50.91 for the NRC’s
processes related to applications to
amend an OL or COL. Section 50.91(b)
stipulates that the Commission will
make available to the licensee the name
of the appropriate State official
designated to receive such amendments.
While the Commission intends to
continue following this practice, the
Commission has not included this
administrative matter in proposed part
53. Proposed § 53.1515(b)(3) contains
some modifications compared to
§ 50.91(b)(3) for clarity; these revisions
are not intended to revise the substance
of the provisions in part 53 compared to
part 50.
Section 53.1520 would be based on
§ 50.92. The section would continue to
use the criteria in § 50.92 for
determining that a proposed
amendment involves no significant
hazards consideration. Although more
specific terms such as event sequence
are used throughout part 53, § 53.1520
would use the term ‘‘accident’’ to
maintain consistency with the long
history of making no significant hazards
consideration determinations under part
50.
Section 53.1525 would provide
requirements for holders of an OL or
COL requesting to revise information
from a DC rule that was referenced in
the initial license application and
included in or incorporated by reference
into the facility FSAR. In keeping with
the current requirements in part 52, the
portion of the part 53 facility licensingbasis information obtained from the
certified design would be divided into
two categories. The most significant
design information and the ITAAC
would be certified by rule and
designated as ‘‘certification
information.’’ The remaining
information, which makes up the
majority of the design information
approved as part of the DC, would not
be certified by rule and is not
considered ‘‘certification information.’’
Part 52 refers to these categories of
information as Tier 1 and Tier 2
information, respectively, and refers to
a change made to that information on a
plant-specific basis as a departure.
Under part 52, a departure from Tier 1
information requires an exemption and,

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for information incorporated into the
license, a license amendment.
Part 53 would dispense with the Tier
1 and Tier 2 terminology. Rather,
§ 53.1525 would use the term
‘‘certification information’’ in place of
Tier 1, and a plant-specific departure
from the certification information would
require both a request for an exemption
from the associated DC rule and, for
information such as ITAAC
incorporated into the license, a license
amendment. However, as would be
provided in § 53.1525(c), a plantspecific departure from the information
approved by the NRC as part of the DC
rule but which is not certification
information (i.e., Tier 2 information
under part 52) would be assessed using
the process and criteria defined in
§ 53.1550 for changes to a FSAR. An
applicant or licensee would need to
identify such a change as a departure
from the referenced standard design in
the updated FSAR. The process for
making a generic change to a certified
design would be described in the
associated section in subpart H.
Section 53.1530 would not allow the
holder of an ML or the holder of a COL
using a manufactured reactor to make
changes to the design of the
manufactured reactor without
requesting a license amendment from
the NRC. This section would provide
the equivalent requirements as those in
§§ 52.98 and 52.171.
Section 53.1535 would establish
requirements for license amendments
during construction. The section would
provide the equivalent options and
requirements for the holders of a CP as
those in § 50.35(b). The regulations
would allow but do not require the
holder of a CP or LWA to request an
amendment under § 53.1510 if the
licensee desires to obtain NRC approval
of a specific design feature or
specification. The requirements for
obtaining an amendment to a COL to
address changes during construction
would also be provided in § 53.1535.
The proposed process would differ from
the current requirements in part 52 by
adopting a requirement that would
explicitly support a change process like
that described in RG 1.237, ‘‘Guidance
for Changes During Construction for
New Nuclear Power Plants Being
Constructed Under a Combined License
Referencing a Certified Design Under 10
CFR part 52.’’
The proposed regulation would allow
the holder of a COL to proceed at its
own risk in making a change during the
construction process and would require
that licensee to submit a license
amendment request no later than 45
days from the date the licensee begins

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to implement the change or departure
requiring NRC approval.
Section 53.1540 would serve as an
introduction to the sections that follow
on changes to licensing-basis
information that are primarily under the
control of a licensee but for which
evaluations are made to determine if a
submittal to the NRC requesting
approval would be required. The section
would also include definitions that
would be applicable when using the
processes in §§ 53.1545 through
53.1565. The definitions would be
largely equivalent to those in § 50.59(a)
but include some revision to reflect the
structure and terminology in other
subparts in part 53. For example, the
definition of ‘‘Change’’ in § 53.1540(b)
would address a ‘‘design feature or
related functional design criteria’’ rather
than a ‘‘design function,’’ because the
former are defined terms in part 53.
Similarly, in § 53.1540(b), the phrase
‘‘design basis’’ from § 50.59(a)(2) would
be replaced with functional design
criteria for SR SSCs.
Section 53.1545 would provide the
proposed requirements for updating of
FSARs. While the process-related
requirements proposed under § 53.1545
would be largely the same as those in
§ 50.71, the specifics of information to
be updated would differ due to the role
of PRA in satisfying the requirements in
subparts B and C. Additionally, the use
of the risk-informed approach in subpart
C would result in some but not all PRA
information being in the FSAR or
another licensing basis document and
therefore a separate PRA update
requirement similar to § 50.71(h) is not
included in proposed subpart I.
Proposed § 53.1239(a)(18) in subpart
H and the related references to this
proposed requirement for the holders of
OLs and COLs would require a
description of the PRA required by
§ 53.450(a) and its results to be included
in FSARs. However, guidance
documents are planned to clarify the
division of PRA-related information that
would need to be in the FSAR, in other
possible licensing basis documents, and
controlled as plant records subject to
inspections and audits. At a minimum,
the information from the PRA that
would be needed to show compliance
with subpart C would be included in the
FSAR (e.g., PRA summary and
analytical results for LBEs). The
submittal of voluminous PRA
information was initially required under
part 52, but that proved to be
impractical and was revised in the 2007
revision of part 52. Guidance is being
developed to ensure sufficient
information is submitted to the NRC to
support the licensing process and the

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NRC’s regulatory findings under part 53
or similar applications using the LMP
under parts 50 or 52.
The NRC has posed a question in
section VI, ‘‘Specific Requests for
Comments,’’ of this document that asks
about the appropriate level of detail for
PRA-related information in an FSAR
and whether other licensing basis
documents might be more appropriate
to both provide information to the NRC
and ensure the PRA is maintained and
updated as proposed in subpart C. The
program document would provide more
detail than the summaries in the FSAR
but still be a much-condensed source of
information in comparison to the
documentation of the PRA.
Section 53.1545(a)(3) and (4) would
be based on the inclusion of at least a
summary of PRA results and the related
margins to safety criteria in the FSAR
and would require updates to that
information. The routine reporting of
these margins would also inform
application of the criteria for allowing
changes without an amendment in the
following section (§ 53.1550) in subpart
I.
Section 53.1550 would establish
requirements for evaluating changes to a
facility as described in its FSAR. This
proposed section would provide the
equivalent of the requirements in
§ 50.59 for evaluating changes to an
FSAR (as updated) and determining if a
license amendment is required to
implement a change to a facility or
procedures. The evaluation criteria
proposed in § 53.1550 would reflect the
role of the PRA in the safety analyses
under part 53 and would include
several measures related to the changes
in plant risk resulting from a change in
the plant design or plant procedures.
Examples would include criteria that
rely on the identification of risksignificant event sequences in
accordance with the analysis
requirements of § 53.450; exceeding the
LBE evaluation criteria as defined in
§ 53.450; the consideration of potential
reductions in margin between the
estimated comprehensive risk metrics
and associated risk performance
objectives in the safety criteria in
§ 53.220; changes to the safety
classification of SSCs; and consideration
of reductions in defense in depth.
Section 53.1550 would include a
criterion related to a departure from a
method of evaluation used in the safety
analyses. The NRC has not yet
developed draft guidance for use in
applying proposed § 53.1550 but
anticipates that the NRC and
stakeholders will assess the potential
need for such guidance and that such
guidance would, if needed, be

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developed as part of ongoing or future
activities.
Section 53.1550 would include
certain concepts taken from existing
guidance for § 50.59 in the proposed
criteria related to DBAs. Specifically,
criterion (iv) for changes made to a
method of evaluation of DBAs under
§ 53.450(f) would be equivalent to a
change in a method of evaluation under
§ 50.59, and criterion (viii) on assessing
if a change creates a possibility for an
accident of a different type than
previously analyzed in the FSAR would
be similar to the § 50.59 criterion (v).
Guidance documents will be prepared
to address the content of applications
for PRA-related information under
proposed part 53, and this guidance will
also influence how potential changes in
the evaluation of LBEs other than DBAs
analyzed under § 53.450(e) are
evaluated and reported under the
proposed criterion (iv).
Section 53.1550(a)(2)(x) would
require evaluating plant changes to
ensure they would not prevent
satisfying the design requirements in
§ 53.440(j) related to the impact of a
large commercial aircraft. The inclusion
of a proposed requirement under
§ 53.1550 related to design features for
protecting against aircraft impact would
reflect the proposed design requirement
in subpart C and related proposed
requirements in subpart H to address
the proposed design requirement in
FSARs.
Sections 53.1560 through 53.1565 in
subpart I would define the processes for
a licensee to evaluate changes to the
program documents included in the
licensing-basis information submitted to
the NRC and to modify such programs
without NRC prior approval.
Section 53.1560 would include the
proposed requirements for updating
program documents included in
licensing-basis information and would
provide the equivalent of FSAR updates
for key program documents. The
proposed requirements in these sections
would provide a uniform approach for
updating program documents, which
correspond to the programs required
under subpart F.
The proposed § 53.1565 would
provide a process for licensees to make
changes to program documents included
in licensing-basis information without
obtaining prior NRC approval. The
proposed requirements would include
several generic criteria that, if not
satisfied, would prompt the need for
NRC approval of a change to a program
document. These generic criteria would
include whether a change would
comply with TS and NRC regulations.
Another proposed criterion for

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evaluating changes to program
documents would be conforming with
program-specific requirements,
including NRC-approved program
documents with more specific criteria
for a particular program, regulations,
administrative controls sections of TS,
and NRC-approved program documents.
Proposed § 53.1565(d) would include
specific criteria for evaluating changes
to several program documents that have
well established change processes and
guidance for licensees under parts 50
and 52. The program documents
specifically addressed in the proposed
section would include quality assurance
programs that would be equivalent to
§ 50.54(a), an emergency preparedness
program that would be equivalent to
§ 50.54(q), and the security program that
would be equivalent to § 50.54(p).
The proposed § 53.1570 would
establish requirements for the transfer of
commercial nuclear plant licenses by
providing the equivalent requirements
of § 50.80 for the possible transfer of an
ESP, CP, OL, or COL. Likewise, the
proposed § 53.1575 would establish
requirements for the termination of an
OL or COL by providing the equivalent
requirements of § 50.82. Other proposed
requirements related to
decommissioning and license
termination would be included in
subpart G.
Section 53.1580 would establish
requirements for information requests
the NRC could send to the various types
of licensees and would provide
requirements that would be equivalent
to requirements in § 50.54(f). The
proposed § 53.1585 would provide the
requirements that would be equivalent
to requirements in § 50.100 to address
revocation, suspension, modification of
licenses, and approvals for cause.
Section 53.1590 would propose to
address backfitting requirements by
providing requirements that would be
equivalent to those in § 50.109.
Proposed § 53.1595 would address
license renewals under part 53 with
simple statements that licenses may be
renewed. This section would be
expanded through future rulemakings to
more fully describe or reference the
processes related to requesting and
processing applications to renew ESPs,
OLs, and COLs issued under part 53 (if
finalized).
Subpart J—Reporting and Other
Administrative Requirements
Part 53 would address various
reporting and administrative
requirements in subpart J.
Section 53.1600 would explain the
organization of the various sections
within the subpart related to providing

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unfettered access to NRC inspectors;
maintaining certain records and
reporting specified events or conditions;
demonstrating compliance with
financial qualification requirements and
providing specified financial reports;
and maintaining financial protections to
address potential accidents.
Section 53.1610 would establish
requirements for the provision of
facilities and unfettered access for
inspections. These requirements would
be equivalent to § 50.70 with only minor
changes proposed to provide additional
flexibilities and address possible
differences related to reactors licensed
under part 53 and the possibility that
some commercial nuclear plants may
not be assigned resident inspectors.
Section 53.1620 would provide for
maintenance of records and the making
of various reports to the NRC. These
requirements would be largely
equivalent to § 50.71. This section is not
intended to reflect all provisions in
§ 50.71; several important requirements
in § 50.71 would be captured in other
sections of part 53. For example,
§ 53.1545 within subpart I would
provide requirements that would be
equivalent to § 50.71(e), updating
FSARs, and § 53.1680, ‘‘Annual
financial reports,’’ would provide the
equivalent of § 50.71(b), which covers
financial reports. A reporting
requirement related to completion of
power ascension testing would be added
to § 53.1620 to support the assessment
of annual fees under 10 CFR part 171,
‘‘Annual Fees for Reactor Licenses and
Materials Licenses, Including Holders of
Certificates of Compliance,
Registrations, and Quality Assurance
Program Approvals and Government
Agencies Licensed by the NRC,’’ which
normally commence upon completion
of those testing activities.
Section 53.1630 would establish
requirements for immediate notification
requirements for operating commercial
nuclear plants. These requirements
would be equivalent to § 50.72 with
minor changes proposed to make the
reporting criteria technology inclusive.
In addition, a new version of NRC Form
361 (NRC Form 361S) would be created
for use by part 53 licensees, but without
LWR-specific terminology to ensure
technology inclusiveness. A separate
rulemaking activity, ‘‘Reporting
Requirements for Nonemergency Events
at Nuclear Power Plants,’’ has been
initiated to consider possible changes to
the requirements in § 50.72. At a future
date, the NRC may consider reconciling
future changes to § 50.72 with the
requirements proposed in part 53,
which have been taken or derived from
the current reporting requirements.

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Section 53.1640 would address the
licensee event report system. These
requirements would be equivalent to
§ 50.73 with minor changes proposed to
make the requirements inclusive of
various reactor technologies and to
reflect appropriate internal references to
other sections in part 53. In addition,
NRC Forms 366, 366A, and 366B would
be revised to include corresponding
check boxes for part 53 licensees.
Section 53.1645 would require
periodic reporting of the quantity of
radionuclides released to unrestricted
areas in liquid and gaseous effluents,
doses to members of the public, and the
results of environmental monitoring.
These reporting requirements in the
proposed part 53 would be largely
equivalent to those in the TSs required
by § 50.36a, ‘‘Technical specifications
on effluents from nuclear power
reactors.’’ The only difference would be
that a § 50.36a requirement to
specifically address conditions where
the dose to the maximally exposed
individual could be significantly above
design objectives would refer to a design
objective of 10 mrem/year total effective
dose equivalent, instead of referring to
the design objectives in appendix I to
part 50. The proposed section would
also include an equivalent to the
reporting requirement in section IV of
appendix I to part 50 if the radiation
exposure to a member of the public in
any calendar quarter exceeds one-half of
the annual ALARA design objective.
Section 53.1650 would include a
reporting requirement to support
safeguards agreements between the
United States and the International
Atomic Energy Agency (IAEA) and
would be equivalent to § 50.78.
Section 53.1660 through 53.1700
would address financial requirements
and would be largely similar to existing
regulations in parts 50 and 52. Section
53.1670 would be entitled ‘‘Financial
qualifications’’ and would require
applicants other than electric utilities to
possess or have reasonable assurance of
obtaining funds for the activities for
which the license is being sought. The
NRC is seeking feedback on these
sections and their ramifications for
merchant plants 3 in section VI,
‘‘Specific Requests for Comments,’’ of
this document. The remaining financial
reports in part 53 would be equivalent
to § 50.71(b) for annual financial
reports, § 50.76 for a change of status,
§ 50.54(cc) for the filing of a petition for
3 A ‘‘merchant plant’’ is a plant licensed to a nonrate-regulated entity (e.g., a nonutility) that engages
in the business of production, manufacturing,
generating, buying, aggregating, marketing, or
brokering electricity for sale at wholesale or for
retail sale to the public.

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bankruptcy, and § 50.81 for creditor
regulations.
Sections 53.1710 through 53.1730
would address financial protection
requirements. Section 53.1720 would
require insurance to stabilize and
decontaminate a plant following an
accident. These requirements would be
taken from § 50.54(w) with the only
notable change being the addition of a
provision allowing plant-specific
estimates of costs to stabilize and
decontaminate a plant as an alternative
to the $1.06 billion minimum coverage
in § 50.54(w). Section 53.1730 is
equivalent to § 50.57(a)(5) and would
refer to the requirements in 10 CFR part
140, ‘‘Financial Protection
Requirements and Indemnity,’’ related
to financial protection requirements and
indemnity agreements, including the
financial protection requirements of the
Price-Anderson Act.

The NRC is proposing a technologyinclusive, risk-informed, and
performance-based approach for the
application of drug and alcohol testing
and fatigue management requirements
for facilities licensed under part 53. The
proposed requirements applicable to
these applicants, licensees, and other
entities would be commensurate with
the radiological consequences presented
by the applicants’ facilities and the
operation of these facilities.4 The

proposed FFD framework would consist
of a two-tiered graded approach similar
to that currently in part 26 and an
optional third tier for part 53
commercial nuclear plants that perform
an analysis that demonstrates the
facility and its operation would satisfy
the criterion in proposed § 26.603(c),
which refers to § 53.860(a). This
proposed FFD framework would be
established in subpart M, ‘‘Fitness for
Duty Programs for Facilities Licensed
Under Part 53,’’ of part 26.
The NRC used operating experience to
provide regulatory flexibility in the
proposed subpart M of part 26
framework to help support a licensee’s
or other entity’s response to changes in
societal drug use, drug testing
technologies and processes, and FFD
program performance. The flexibility
would also help in FFD program
implementation because of the wide
variety of staff sizes anticipated at
commercial nuclear plants licensed
under part 53 and the geographically
remote locations in which commercial
nuclear plants may be sited.
The proposed first-tier FFD program
requirements would apply to part 53
licensees and other entities of
commercial nuclear plants under
construction who satisfy the criterion in
§ 26.603(c) but elect not to implement
proposed § 26.604, ‘‘FFD program
requirements for facilities that satisfy
the § 26.603(c) criterion,’’ or who do not
satisfy the criterion in § 26.603(c), and
to holders of MLs who are assembling
or testing manufactured reactors. These
requirements would be provided in
proposed § 26.605(a) and would be
essentially equivalent to those
requirements in subpart K, ‘‘FFD
Program for Construction,’’ of part 26 as
supplemented by select requirements
from subparts E, ‘‘Collecting Specimens
for Testing,’’ and I, ‘‘Managing Fatigue,’’
of part 26, and the requirements in
subparts A, ‘‘Administrative
Provisions,’’ and O, ‘‘Inspection,
Violations, and Penalties,’’ of part 26.
The first-tier requirements would
involve policies, procedures, behavioral
observation, fatigue management, drug
and alcohol testing, determinations of
fitness, appeals, training, sanctions,
auditing, change control, performance
monitoring, recordkeeping, and
reporting. These requirements would
help deter individuals subject to this
section from illicit drug and/or alcohol
use and from being impaired from any
cause including fatigue. These proposed
requirements would also help licensees

4 The NRC uses the term ‘‘operation’’ in its part
26 discussion to focus on human performance,
namely the necessity of individuals to operate,

maintain, surveil, and protect the facility and
respond to operational transients and unlikely
event sequences.

Subpart M—Enforcement
Subpart M would contain two
provisions, § 53.9000 and § 53.9010,
which are analogous to provisions
contained in other parts of 10 CFR
Chapter I imposing requirements on
regulated entities. Section 53.9000
would provide notice of the
Commission’s authority under the Act
to obtain injunctions or other court
orders for the enumerated violations.
Paragraph (a) of § 53.9010 would
provide notice to all persons and
entities subject to part 53 that they are
subject to criminal sanctions for willful
violations, attempted violations, or
conspiracy to violate certain regulations
under part 53. Criminal sanctions would
not apply to the regulations listed in
paragraph (b). The regulations for which
criminal penalties would apply are
limited to those that establish either a
regulatory obligation or prohibition.
V. Changes to Other Parts of 10 CFR
Chapter I
10 CFR Part 26
A. Introduction

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and other entities identify individuals
as users of impairing substances and
demonstrate compliance with § 26.23,
‘‘Performance objectives.’’
The proposed second tier would
include all the proposed first-tier
requirements, plus the more
comprehensive set of FFD program
requirements in current subparts C,
‘‘Granting and Maintaining
Authorization,’’ D, ‘‘Management
Actions and Sanctions to be Imposed,’’
H, ‘‘Determining Fitness-for-Duty Policy
Violations and Determining Fitness,’’
and N, ‘‘Recordkeeping and Reporting
Requirements,’’ of part 26. These
requirements would be provided in
proposed § 26.605(b) and would be
applicable to licensees and other
entities satisfying the § 26.603(c)
criterion, at their discretion. These
requirements would also apply to
licensees or other entities not satisfying
the § 26.603(c) criterion that implement
an FFD program under subpart M of part
26, before the loading of fuel onsite into
a reactor vessel; before receiving a
manufactured reactor; or before
operating, testing, performing
maintenance of, or directing the
maintenance or surveillance of securityrelated equipment or equipment that a
risk-informed evaluation process has
shown to be significant to public health
and safety.
The second-tier requirements are
based on the additional risk presented
by nuclear reactor assembly, testing,
fueling, and operation and the necessity
for human actions in certain event
sequences. The inclusion of the current
part 26 requirements would align
proposed part 53 FFD and AA program
requirements with the current FFD and
AA programs required for facilities
licensed under parts 50 and 52. This
approach would ensure effective and
consistent AA and FFD program
implementation across the commercial
nuclear power industry, thereby
ensuring uniform requirements for
individuals who may perform roles and
responsibilities for multiple facilities
regardless of facility licensure.
Proposed § 26.604 would offer an
alternate option for an applicant
implementing an FFD program under
subpart M of part 26. If the applicant
demonstrates that the criterion in
proposed § 26.603(c) is met, then the
applicant (and the subsequent licensee
or other entity) must still implement an
FFD program described in subpart M of
part 26; however, drug and alcohol
testing would not be required unless
FFD performance declines or the
applicant, licensee, or other entity elects
to implement drug and alcohol testing.
The proposed § 26.604 requirements are

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equivalent to those proposed in
§ 26.605(a) except for required drug and
alcohol testing. This proposed
framework would focus on the human
performance of individuals while they
are performing those duties and
responsibilities that make them subject
to the FFD program. This performance
would be verified through behavioral
observation, evaluation of any FFD
concerns, performance monitoring,
fatigue management, and
determinations of fitness. Applicants
that do not satisfy the criterion in
proposed § 26.603(c), or elect not to
perform the analysis required to
demonstrate that the criterion in
§ 26.603(c) is met, would be subject to
an FFD program described in § 26.605,
‘‘FFD program requirements for
facilities that do not implement
§ 26.604,’’ or an FFD program that
implements all part 26 requirements,
except for those requirements in
subparts K and M of part 26.
In establishing the minimum FFD
program requirements in § 26.604, the
NRC reviewed current advanced reactor
designs against that of a non-power
production or utilization facility (NPUF)
that is not required to implement an
FFD program for those individuals who
have unescorted access to the controlled
access area (and vital area for some
facilities), including NRC-licensed
operators.5 This review was performed
because commercial nuclear plants
licensed under part 53 could be
designed with similar power levels and
radiological consequences as the
currently licensed NPUFs. From this
review, three principal considerations
supported the minimum set of
requirements for the § 26.604 FFD
program.
First, the radiological consequences
presented by a part 53 licensed facility
and its operation that satisfy the
criterion in § 26.603(c) may present a
greater potential radiological
consequence to workers and the public
in the vicinity of the facility than does
an NPUF. Second, the operating
characteristics of a part 53 licensed
facility are unlike that of an NPUF
because there may be a higher reliance
on individuals at the part 53 site to
safely and competently operate,
maintain, surveil, and secure SSCs that
may not be required at an NPUF, such
as systems that provide secondary heat
transfer, reactor coolant flow, pressure
control, and at-power core refueling.
Differences in operating characteristics
could include, for example: long-term,
full power operation with automated
5 Controlled access area and vital area are defined
in § 73.2, ‘‘Definitions.’’

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reactivity control systems for loadfollowing; active and passive safety and
security systems; innovative non-lightwater heat transfer systems; and energy
storage and hazardous chemical
systems. The individuals at part 53
facilities may also be required to
communicate to individuals both onsite
and offsite, such as electrical load
dispatchers, any conditions adverse to
safety, security, or quality. Third, part
53 licensed facilities may be sited in
geographically remote locations that
may not have a physically available
administrative or corporate support
team to provide face-to-face oversight,
engineering expertise, and maintenance
support like that at NPUFs. This places
a higher reliance on those individuals
required at a part 53 facility being fit for
duty and trustworthy and reliable
because a replacement individual may
not be readily available.
The NRC proposes to exclude drug
and alcohol testing from the proposed
§ 26.604 framework for five reasons: (1)
the § 26.23 performance objectives can
be met through effective
implementation of the defense-in-depth
regulatory framework established by
behavioral observation, reporting of
legal actions, the proposed performance
monitoring and review program (PMRP),
FFD training, and requirements from the
physical protection, AA, cyber
protection, and licensed operator
programs; (2) the PMRP would require
the licensee or other entity to monitor
its FFD program performance (both
qualitatively and quantitatively) against
its historical site performance, fleetlevel performance, if applicable, and
industry performance. The licensee or
other entity would be required to
implement corrective actions if site FFD
performance meets a licensee- or other
entity-established threshold or to
resolve a finding resulting from a
qualitative review or audit in a manner
that restores performance and corrects
root causes, contributing causes, or
both; (3) the requirements in proposed
§ 26.609, ‘‘Behavioral observation,’’ are
more robust than those in § 26.407,
‘‘Behavioral observation,’’ of subpart K
of part 26 and are proposed to
synchronize with and reinforce the AA
behavioral observation requirements in
§ 73.56, ‘‘Personnel access authorization
requirements for nuclear power plants,’’
or the proposed requirements under
§ 73.120, ‘‘Access authorization program
for commercial nuclear plants’’; (4) a
part 53 commercial nuclear plant that
satisfies the § 26.603(c) criterion will be
designed, operated, and secured with a
radiological risk profile that is lower
than that described in § 53.860(a)(2) and

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perhaps will approach the radiological
risk profile of an NPUF (which does not
implement an FFD program); and (5) the
NRC is aware that a part 53 commercial
nuclear plant could be designed and
constructed in such a manner to reduce
reliance on an onsite security force to
protect SSCs, NRC-licensed materials,
and sensitive information, with
enhanced capabilities for the detection,
assessment, and delay of a DBT
adversary.
Regarding fatigue management
requirements, work hour controls would
be required for personnel at utilization
and manufacturing facilities in
accordance with the existing scoping
criteria in § 26.4, ‘‘FFD program
applicability to categories of
individuals,’’ as revised in this
proposed rule. The amended § 26.4 also
would be used to determine whether an
individual would be subject to drug and
alcohol testing. The applicability of
these scoping criteria for certain
individuals (such as operators and
maintenance personnel) would be
determined by the licensee or other
entity through its risk-informed
evaluation process performed to assess
the risk significance of the SSC upon
which work is being performed or
directed by the individual. These
requirements also would be scaled
based on the potential radiological
consequences presented by the facility.
However, fatigue management would be
applied to all individuals subject to the
FFD program, similar to FFD program
implementation by the current fleet of
commercial nuclear plants because
fatigue management is a proactive
requirement designed to help prevent
on-shift impairment through work hour
scheduling and time off. The behavioral
observation program (BOP) would be
the principal requirement to provide
reasonable assurance that individuals
on shift are not mentally or physically
impaired due to fatigue, which in any
way could adversely affect their ability
to safely and competently perform their
duties.
The NRC is proposing subpart M of
part 26 for facilities licensed under part
53, in lieu of subjecting all part 53
licensees to the same part 26
requirements that apply to facilities
licensed under part 50 or 52, for four
principal reasons. First, subpart M of
part 26 would apply FFD requirements
in a risk-informed manner
commensurate with the radiological
consequences presented by facilities
licensed under part 53. This regulatory
strategy is consistent with the current
part 26, which provides a
comprehensive set of deterministic
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entities at facilities that are operating.
This approach is also consistent with
the current subpart K of part 26, which
provides a more flexible framework for
nuclear power reactors under
construction, where the probabilities of
serious radiological accidents are lower
and consequences from such accidents
are less severe than at operating plants.
Second, subpart M of part 26 would
enable a part 53 licensee or other entity
to implement innovative drug testing
technologies and behavior observation
techniques while continuing to
demonstrate compliance with the part
26 performance objective in § 26.23(b) of
providing reasonable assurance that
individuals are not under the influence
of any substance or mentally or
physically impaired from any cause,
which in any way adversely affects their
ability to safely and competently
perform assigned duties. These
technologies include drug and alcohol
testing using oral fluid, urine, and hair
specimens; screening using point of
collection testing and assessment
(POCTA) devices; and monitoring using
passive drug and alcohol detection
instrumentation. Part of the basis to
enable the use of innovative drug and
alcohol testing technologies is to
maintain FFD program effectiveness
should the staff size at a part 53
commercial nuclear plant be small and
challenge the effective implementation
of the behavioral observation and drug
and alcohol testing programs. Also, a
commercial nuclear plant that is sited at
a geographically remote location may
present additional challenges to
behavioral observation and drug and
alcohol testing that are not presented by
traditional LWR facilities licensed
under part 50 or 52, such as: efficiency
of postal services for shipping and
controlling biological specimens;
proximity to drug and alcohol collection
facilities that are reasonably equivalent
to that described in subpart E of part 26;
availability of internet and cellular
services to enable same-time
discussions among the Medical Review
Officer (MRO), donor, and laboratory;
accessibility to substance abuse
treatment services described in subpart
H of part 26; and proximity to an MRO
(or management and clinical staff) to
evaluate potential impairment caused
by fatigue and/or substance use or
abuse, for-cause and post-event
occurrences, and the individual’s
potential to return to duty.
A part 53 commercial nuclear plant
that is sited in a geographically remote
location and has a small staff size may
present implementation challenges and
the potential for small group dynamics
to impact FFD program effectiveness.

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Particularly in isolated environments,
psychological phenomena known as
‘‘groupthink’’ may take effect and could
impact the effectiveness of BOPs and
the ability to effectively manage safety
culture. For example, in circumstances
where small staffs are drawn from the
same small town and thereby have a
potentially narrow experience base, it
could be challenging to maintain a
safety conscious work environment in
which personnel feel free to raise safety
concerns without fear of retaliation,
intimidation, harassment, or
discrimination, and organizations may
resultingly experience groupthink-like
effects. Groupthink is particularly
prevalent among cohesive and insulated
groups that experience high levels of
decisional stress.6 Small staffs at part 53
commercial nuclear plants may
therefore be more susceptible to
groupthink if they are working in an
isolated environment where decisionmaking pressures may be high.
Groupthink could have adverse effects
on workplace safety culture, as studies
show that individuals will be more
hesitant to speak out against practices
they deem unsafe for fear of deviating
from group norms.7 Individuals may
also be unaware of systematic biases in
the group decision-making process and
may then be less likely to scrutinize the
potential risks of the group’s decision or
sufficiently contemplate alternative
paths of action.8 Furthermore, the
literature indicates that groups make
riskier decisions than individuals acting
alone due to the diffusion of
responsibility among group members.9
6 See e.g., Irene W#r<, Ragnar Rosness, and Stine
Skaufel Kilska, ‘‘Human performance and safety in
Arctic environments,’’ SINTEF (2018).
7 See e.g., Russell Mannion and Carl Thompson,
‘‘Systematic biases in group decision-making:
implications for patient safety,’’ International
Journal for Quality I Health Care, Vol. 26, No. 6
(2014): 606–612 (arguing that small group dynamics
in healthcare teams produce systematic biases in
group decision-making because healthcare
professionals may be reticent to vocalize concerns
they have about quality of care).
8 See e.g., W#r<, Rosness, and Kilska (arguing
that groupthink leads teams to ‘‘develop shared
rationalizations that bolster a proposed choice,
rather than examining alternative options and
identifying the risks associated with the proposed
choice’’). See also David Hofmann and Adam
Stetzer, ‘‘A Cross-Level Investigation of Factors
Influencing Unsafe Behaviors and Accidents,’’
Personnel Psychology, Vol. 49 (1996) (finding that
in a study of fatal accidents involving offshore oil
rigs, in the absence of standard operating
procedures, workers ‘‘equated normal work
methods (i.e. what everyone else does) with safe
and/or ideal work methods,’’ revealing that the
groupthink phenomena will further cement modes
of work that do not reflect safety protocols in small
groups that lack strong norms around workplace
safety and tacitly reward short-cuts that prioritize
efficiency over safety).
9 Mannion and Thompson, ‘‘Systematic biases in
group decision-making: implications for patient

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This phenomenon, known as ‘‘the risky
shift,’’ also runs counter to a safety
culture. Accordingly, ‘‘groupthink’’ and
‘‘the risky shift’’ may lead to group
behaviors that render behavioral
observation less effective. As such,
alternative approaches to behavior
observation programs, such as the
utilization of video-based surveillance
by individuals separate from the onsite
work unit, could serve to mitigate
potential issues associated with
groupthink. The incorporation of remote
observation, performed by individuals
physically separate from the site, could
help to bring in independent and
objective perspectives and help to break
patterns of thought and communication
that may result in groupthink.
Even without the influence of small
group dynamics, there are other
practical constraints to implementing
FFD requirements, such as random drug
and alcohol testing, among small staffs.
Random testing is less effective when
applied to small staff sizes because it
may be easier for staff to communicate
and predict when individuals will be
subject to drug and alcohol testing.
Furthermore, if a facility is sited in a
remote location, program
implementation could be challenged by
the following factors: limited mail
services to laboratories certified by the
U.S. Department of Health and Human
Services (HHS), availability of local
clinical or medical options for treatment
and determinations of fitness by an
MRO or Substance Abuse Expert, and
use of offsite drug and alcohol
collection facilities.
The increased potential for small staff
sizes to impact FFD policy compliance
warrants an approach to FFD that
emphasizes performance over
prescriptive requirements that may be
ineffective or infeasible at these
facilities. Therefore, the NRC proposes
the subpart M of part 26 framework to
provide a performance-based approach
to FFD. For example, proposed
§ 26.603(d) would use existing part 26
auditing requirements and the reporting
requirement in § 26.717, ‘‘Fitness-forduty program performance data,’’ and
clarify how FFD performance data
would be used to maintain or improve,
if necessary, FFD program effectiveness.
Specifically, § 26.603(d) would require
each licensee and other entity that elects
to implement subpart M of part 26 to
monitor and assess their site-specific
performance against the preceding
year’s site performance, the licensee’s
most recent fleet-level performance, and
the most recent industry performance.
safety,’’ International Journal for Quality I Health
Care, Vol. 26, No. 6 (2014): 606–612.

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Licensees and other entities would use
these datapoints to develop performance
measures, which would be qualitative
descriptions of the specific FFD
program elements, and threshold values
for each performance measure that, if
exceeded, would indicate a performance
deficiency. Each licensee and other
entity would compare its site’s current
performance data against the
performance measures and, if a
threshold is exceeded, the licensee or
other entity would be required to take
corrective actions to restore
performance. Also, the NRC proposes a
change control requirement to allow a
licensee or other entity to change its
subpart M of part 26 FFD program while
ensuring that FFD program effectiveness
is maintained.
Lastly, subpart M of part 26 would
consolidate the applicable FFD
requirements by placing in one subpart
all proposed part 26 requirements
(either new requirements or crossreferences to existing part 26
requirements) for part 53 licensees and
other entities. This should help
licensees and other entities implement
the requirements because it would
enable easy cross-reference to similar
requirements in other subparts that are
being implemented by non-part 53
licensees and entities subject to part 26.
Understanding how other licensees or
other entities implement similar FFD
requirements may facilitate the sharing
of operating experience in program
implementation.
The use of innovative technologies
and a risk-informed performance-based
framework parallels the considerations
presented in the Advanced Reactor
Policy Statement. As stated in the policy
statement, ‘‘[S]implified systems should
facilitate operator comprehension,
reliable system function, and more
straightforward engineering analysis.’’
Furthermore, these same attributes may
reduce potential radiation exposures,
help prevent the theft of nuclear
materials, and use technology and
design innovations. Should these
components and systems be designed,
implemented, and maintained to
minimize reliance on human actions
and leverage technology and innovation,
then the robust and prescriptive FFD
requirements in, for example, subparts
B, ‘‘Program Elements,’’ and E of part 26
could be scaled to the part 53-licensed
facility and its operation. This strategy
would be implemented in the subpart M
of part 26 framework.
Even though current subpart K of part
26, provides for FFD requirements
commensurate with the radiological
consequences presented by a nuclear
power plant construction site, proposed

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subpart M of part 26 would not allow
part 53 licensees and other entities to
implement the requirements in subpart
K. The principal reasons are that
(without significant changes to subpart
K that would be outside the scope of
this rulemaking): (1) subpart K does not
apply to holders of MLs who assemble
or test a reactor; (2) subpart K only
applies during construction, whereas
subpart M would apply during
construction, operation, and
decommissioning through
implementation of the insider
mitigation program (IMP) required by
§ 73.55 or proposed § 73.100; (3) subpart
K does not address training,
authorization as defined in § 26.5, and
MRO performance; (4) subpart K does
not expressly authorize the use of
innovative drug and alcohol testing
technologies; (5) subpart K does not
describe the use of time-dependent
alcohol limits or special analysis testing
of dilute urine specimens; and (6)
subpart K has less rigor in the protection
of worker rights and sensitive
information than that proposed in
subpart M.
Despite the differences between
subparts K and M of part 26, the
requirements in subpart M would be
essentially equivalent to many in
subpart K that were implemented by the
licensees of Vogtle Nuclear Station and
V.C. Summer Nuclear Station when they
were constructing four commercial
nuclear power reactors and NRC
inspection and operating experience
evaluation determined that the use of
subpart K contributed to adequately
protecting the public health and safety
and the common defense and security.
Further, given the risk profile posed by
facilities licensed under part 53 and the
proposed additional requirements in
subpart M of part 26 that were
developed from operating experience
and other part 26 subparts (but are not
included in subpart K of part 26), the
NRC concludes that if licensees and
other entities effectively implement the
proposed requirements in subpart M of
part 26, then individuals subject to the
rule should be fit for duty and
trustworthy and reliable.
B. Proposed Changes to Part 26,
Subparts A Through E and I
Section 26.3(d) is the applicability
paragraph for contractor/vendors (C/Vs)
who implement FFD programs or
program elements, to the extent that the
licensees and other entities specified in
§ 26.3(a) through (c) rely on those C/V
FFD programs or program elements to
meet the requirements of part 26.
Section 26.3(d) would be amended to

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address part 53 licensees and other
entities in proposed § 26.3(f).
Proposed § 26.3(f) would place part 53
licensees or other entities within the
scope of part 26. For licensees and other
entities of a part 53 commercial nuclear
plant, except a holder of an ML, the FFD
program would be required to be
implemented no later than the start of
construction activities. The holder of an
ML would need to implement its FFD
program before commencing activities
that assemble a reactor.
Current § 26.4 describes FFD program
applicability to categories of
individuals. These categories are based
on the duties, responsibilities, and the
types of access an individual may
possess. The NRC proposes to amend
§ 26.4 to include licensees and other
entities described in § 26.3(f). The NRC
expects that not all categories of
individuals described in current § 26.4
would be applicable to all part 53
facilities. The NRC is proposing
regulatory guidance in DG–5073,
‘‘Fitness-of-Duty Programs for
Commercial Nuclear Plants and
Manufacturing Facilities Licensed
Under 10 CFR part 53,’’ and DG–5078,
‘‘Fatigue Management for Nuclear
Power Plant Personnel at Commercial
Nuclear Plants Licensed Under 10 CFR
part 53,’’ to help address program
applicability to certain individuals.
Section 26.4(a)(1) and (a)(4) would be
amended to account for the possibility
that certain individuals may perform or
direct the performance of operational
and maintenance activities from a
remote facility (for example, a remotecontrol station) for licensees or other
entities licensed under part 53.
The framework of the current part 26
does not account for individuals who
perform operating and maintenance
duties at remote facilities. Although
current § 26.4(a)(1) does not limit the
operating of applicable SSCs to onsite
operating, § 26.5 limits the definition of
‘‘Maintenance,’’ for the purposes of
§ 26.4(a)(4), to include only ‘‘onsite
maintenance activities.’’ In the 2008
part 26 final rule preamble, the NRC
explained that the work hour
requirements apply to those individuals
who perform maintenance activities
within the licensee’s owner-controlled
area. Furthermore, regarding the
direction of applicable operations and
maintenance activities, current
§ 26.4(a)(1) and (4) address only
individuals who perform ‘‘onsite
direction.’’
Under the proposed amendments to
part 26, the limitation of ‘‘onsite’’
activities to those performed within the
owner-controlled area would still apply
to facilities licensed under part 50 or 52.

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However, for licensees and other
entities described in § 26.3(f), the NRC
would remove the ‘‘onsite’’ limitation to
include activities performed both within
the owner-controlled area as well as
operations and maintenance duties
performed at remote facilities where
safety-significant systems and
components are expected to be operated
within the design basis of the
commercial nuclear plant.
In the 2008 part 26 final rule, the
purpose of limiting ‘‘directing’’
activities to those ‘‘directing’’ activities
that are conducted onsite was to avoid
requiring work hour controls for
individuals performing incidental
duties, consistent with § 26.205(b)(5),
from an offsite location in instances
where those duties might be considered
to be ‘‘directive’’ in nature. Under the
proposed amendments to part 26, the
exclusion of incidental duties while
calculating work hours would still be
applicable for licensees and other
entities licensed under part 53.
However, for these licensees and other
entities, beyond instances of incidental
duties, the direction of operations and
maintenance activities associated with
safety-significant SSCs, when performed
at remote facilities, would be considered
in an equivalent fashion as direction
performed at non-remote facilities, for
the purposes of administering work
hour controls.
Proposed § 26.4(b) would include in
an FFD program individuals who are
granted unescorted access to the
protected area of a facility licensed
under part 53 and do not perform or
direct the performance of the duties
described in § 26.4(a). This requirement
would contribute to the defense-indepth regulatory framework that helps
provide that individuals who have
unescorted access are fit for duty,
trustworthy, and reliable. For example,
the NRC is proposing amendments to
part 73 to require a part 53 licensee to
subject individuals to a series of reviews
to help determine whether those
individuals are trustworthy and reliable
before granting them unescorted access
to the facility’s protected area.
The NRC would amend § 26.4(c) to
include in an FFD program individuals
who are assigned to physically report to
the part 53 licensee’s emergency
response facility (or facilities) or
participate remotely in emergency
response activities, and individuals
without unescorted access to the part 53
facility who, remotely or otherwise,
make decisions and/or direct actions
regarding plant safety or security. Part
53 commercial nuclear plants may be
licensed for and rely upon offsite
facilities to fulfill the role of a Technical

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Support Center or Emergency
Operations Facility. Therefore, the
proposed rule would account for such
offsite facilities or remotely performed
activities. Further, the use of personnel
to operate systems and components,
maintain and surveil SSCs, and respond
to plant conditions and security events
may be different than those included in
the Technical Support Center or
Emergency Operations Facility team for
power reactors currently licensed under
part 50 or part 52.
For the individuals whose duties for
the licensees and other entities in
§ 26.3(c) require the individuals to have
the types of access or perform the
activities listed in § 26.4(e)(1) through
(6) at the location where the commercial
nuclear plant will be constructed and
operated, current § 26.4(e) requires them
to be subject to an FFD program that
satisfies all the requirements of part 26
except subparts I and K. The NRC
would amend § 26.4(e) to except subpart
M as well as subparts I and K. The NRC
would also amend § 26.4(e) to include
in an FFD program the individuals
whose duties for the licensees and other
entities in § 26.3(f) require the
individuals to have the types of access
or perform the activities listed in
§ 26.4(e)(1) through (6) or perform
construction activities as defined in
§ 26.5.
Section 26.4(e)(4) would be revised to
include in an FFD program individuals
who witness or determine inspections,
tests, and analyses certifications
required under part 53 because current
§ 26.4(e)(4) includes the individuals
who perform the same duties under part
52.
The proposed rule would amend
§ 26.4(f) to require individuals who
construct or direct the construction of
safety- or security-related SSCs at
facilities licensed under part 53 to be
subject to an FFD program under
subpart M of part 26 or an FFD program
that demonstrates compliance with all
of the requirements of part 26 except for
subparts I, K, and M of part 26.
Section 26.4(g) is the applicability
paragraph for FFD program personnel
(e.g., the FFD manager, MRO, and
technicians) and persons who perform
AA determinations (e.g., the licensee- or
other entity-designated Reviewing
Official). This section would be
amended to address part 53 licensed
facilities. Specifically, a part 53 licensee
or other entity would use FFD program
personnel to implement its FFD
program as well as other assigned
individuals who are not involved in the
day-to-day operations of the program to
implement specific elements of its FFD
program, such as the collection of a

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specimen for drug or alcohol testing.
These individuals would be held
accountable for program
implementation, including consistent
implementation of protections afforded
to all individuals subject to the FFD
program.
Section 26.4(h) would be amended to
include subpart M of part 26.
The NRC proposes to include several
new definitions in § 26.5, ‘‘Definitions,’’
and amend some existing definitions.
The NRC is proposing to add a
definition for ‘‘Biological marker.’’ The
proposed definition would be consistent
with ‘‘Biomarker’’ defined by the HHS
in its Mandatory Guidelines for Federal
Workplace Drug Testing (HHS
Guidelines) using oral fluid as the
biological specimen to be tested (84 FR
57554; October 25, 2019). However, the
proposed definition for § 26.5 would
add that the endogenous substance used
to validate that the biological specimen
‘‘was produced by the donor’’ because
subpart M of part 26 proposes to have
the MRO evaluate any discrepant
biological marker identified in a
biological specimen collected from a
donor.
The NRC is proposing a definition for
the word ‘‘Change’’ as used in the
proposed § 26.603(e), ‘‘FFD program
change control,’’ process. The proposed
definition would be consistent with the
definition of ‘‘Change’’ for a part 50 or
52 licensee’s emergency plans in
§ 50.54(q)(1)(i).
The NRC proposes to revise the
definition of ‘‘Constructing or
construction activities’’ to clarify that
for licensees or other entities in
§ 26.3(f), the definition of
‘‘Construction’’ would be that as
proposed in § 53.020.
The definitions of ‘‘Contractor/
vendor’’ (C/V) and ‘‘Other entity’’ would
be revised to make them applicable to
part 53 licensees. A holder of an ML
under part 53 could be a C/V under the
proposed C/V definition.
The NRC is proposing a definition for
‘‘Illicit substance’’ because this phrase is
used in subpart M of part 26 and would
address substances that cause
impairment and possible addiction but
are not an ‘‘illegal drug’’ as defined in
§ 26.5. This proposal is based on
operating experience where individuals
have admitted to using common
household, non-drug substances to
achieve a high or satisfy an addiction.
These common household items
include, but are not limited to nitrous
oxide, butane, propane, glue, paint
vapors, lighter fluid, nail polish
remover, degreasers, permanent
markers, and methyl alcohol (which is

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found in hand sanitizer and
mouthwash).
The definition of ‘‘Questionable
validity’’ would be revised to make it
applicable to an FFD program
implemented under subpart M of part
26, which would include all biological
specimens.
The NRC is proposing a definition for
‘‘Reduction in FFD program
effectiveness’’ because this phrase,
similar to the proposed definition for
‘‘Change,’’ is used in proposed
§ 26.603(e). The proposed definition is
generally consistent with the definition
of ‘‘Reduction in effectiveness’’
provided for emergency plans in
§ 50.54(q)(1)(iv).
The proposed rule would make the
current definition of ‘‘Reviewing
official’’ applicable to those licenses and
other entities in § 26.3(f).
The current part 26 definition of
‘‘Safety-related structures, systems, and
components’’ would be amended to use
the NRC’s proposed definition in
§ 53.020 for the part 53 licensees and
other entities described in § 26.3(d) and
(f).
The NRC would amend the definition
of ‘‘Security-related SSCs’’ in § 26.5 to
make it applicable to a licensee or other
entity described in § 26.3(d) and (f).
The NRC proposes a definition for
‘‘Special Nuclear Material’’ that would
refer to the definition in § 70.4,
‘‘Definitions,’’ of part 70 to ensure
consistency.
The NRC is proposing a revision of
the definition of ‘‘Unit outage’’ to
account for the potential use of
commercial nuclear plants for purposes
other than electricity generation.
Section 26.21, an applicability
statement for part 26 FFD programs,
would be amended to include licensees
and other entities described in § 26.3(f)
that choose to implement an FFD
program that implements all part 26
requirements, except those in subparts
K and M of part 26.
Section 26.51, ‘‘Applicability,’’ would
be amended to apply to licensees and
other entities described § 26.3(f) that
elect not to implement the requirements
in subpart M of part 26 for the categories
of individuals in § 26.4 and those
licensees and other entities that elect to
implement the requirements in § 26.605.
Section 26.53(e), (e)(1) and (3), and (g)
through (i), which are general
provisions for granting and maintaining
authorization, would be amended to
apply to licensees and other entities
described § 26.3(f).
Section 26.63(d), a suitable inquiry
requirement, would be amended to
apply to licensees and other entities
described § 26.3(f).

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Section 26.73, the applicability
statement for subpart D of part 26,
would be amended to apply to licensees
and other entities described § 26.3(f)
that elect not to implement the
requirements in subpart M of part 26 for
the categories of individuals in § 26.4
and those licensees and other entities
that elect to implement the
requirements in § 26.605(b).
Section 26.81, the purpose and
applicability statement for subpart E of
part 26, would be amended to apply to
licensees and other entities described in
§ 26.3(f) that elect not to implement the
requirements in subpart M of part 26 for
the categories of individuals in § 26.4
and those licensees and other entities
that implement proposed § 26.605(a) or
(b). The subpart E requirements to be
implemented are listed in proposed
§ 26.607(c)(2)(i) and (c)(2)(ii) and (c)(3).
Section 26.201, the applicability
statement for subpart I of part 26 would
be amended to apply to licensees and
other entities described in § 26.3(f).
Also, the applicability statement would
be divided into two paragraphs for
clarity.
The NRC proposes to add § 26.202,
‘‘General provisions for facilities
licensed under part 53,’’ for licensees or
other entities described in proposed
§ 26.3(f) that elect to implement the
requirements in subpart I of part 26 in
accordance with § 26.604 and § 26.605.
Section 26.202 would establish
requirements equivalent to those in
current § 26.203, ‘‘General provisions,’’
which is applicable to part 50 and 52
licensees. The NRC would add the
separate § 26.202 because § 26.203 refers
to various requirements under subpart B
of part 26, which would not be
applicable to facilities licensed under
part 53 that implement subpart M of
part 26.
Additionally, § 26.202(c), ‘‘Training
and assessments,’’ unlike § 26.203(c),
‘‘Training and examinations,’’ would
not include a comprehensive
examination requirement because
trainee assessment is conducted as part
of a SAT that would be required as
proposed under the FFD program
training requirements in § 26.608.
Proposed changes in §§ 26.205,
26.207, and 26.211 would add
references to new requirements in
subparts I and M of part 26 that would
be applicable specifically to licensees
and other entities in § 26.3(f). The NRC
would not change the specific
provisions for work hour requirements
in current § 26.205(d). However, as
addressed in the discussion of proposed
changes to § 26.4(a), whether a licensee
or other entity under part 26 would
need to implement work hour controls

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for certain individuals or groups would
be dependent, in part, on
determinations reached by that
licensee’s risk-informed evaluation
process.
Proposed changes to
§§ 26.207(a)(1)(ii) and 26.211(b) would
allow licensees and other entities in
§ 26.3(f) to perform face-to-face
assessments to support the approval of
work hour control waivers and the
conduct of fatigue assessments,
respectively, using electronic
communications. These proposals
would allow supervisors to conduct
such assessments from a remote location
under appropriate circumstances. Such
remotely conducted assessments would
need to be supported by someone who
is present in-person with the individual
being assessed and who is trained in
accordance with the requirements of
either § 26.29 and § 26.203(c) or § 26.608
and § 26.202(c). The reasoning for these
proposals and the associated need for
in-person support to augment electronic
communications is addressed further in
the preamble discussion of proposed
§ 26.619.
C. Proposed Requirements for Part 26,
Subpart M
The proposed rule would add a new
subpart M to part 26 that would provide
alternative FFD requirements for part 53
licensees and other entities.
Proposed § 26.601 would make
subpart M of part 26 applicable to part
53 licensees and other entities, at their
discretion. If a licensee or other entity
in § 26.3(f) does not elect to implement
an FFD program that demonstrates
compliance with the requirements of
subpart M, then the individuals
specified in § 26.4 would be subject to
an FFD program that demonstrates
compliance with all part 26
requirements, except for those
requirements in subparts K and M.
Proposed § 26.603(a) would require an
applicant to provide a description of its
FFD program and its implementation
within its application for a license. This
requirement is equivalent to the existing
requirements in §§ 26.401(b) and
52.79(a)(44). The entities that would be
required to submit these FFD program
descriptions are certain applicants that
would comply with the part 53
application requirements in subpart H.
In subpart H, § 53.1309(a)(6) would
require an applicant for a CP to provide
a description of its FFD program in its
PSAR. Under §§ 53.1279(b)(4),
53.1369(x), and 53.1416(a)(24), an
applicant for an ML, OL, and COL,
respectively, would be required to
provide a description of its FFD
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Unlike an application for a license, a
description of an FFD program does not
receive NRC review for possible
approval. The applicant provides the
NRC with information about the
applicant’s proposed FFD program to
inform the NRC’s inspection program
and to demonstrate that the FFD
program will be effectively
implemented before a licensee or other
entity commences any activity making
individuals at the NRC-licensed facility
subject to the FFD program.
Proposed § 26.603(a)(1) would require
a summary description of the analysis
described in § 26.603(c), if performed.
The analysis should describe the
operation of the facility. This would
include informing the Commission of:
(1) the principal individuals assigned by
job title (work category) and a summary
description of the human actions (e.g.,
monitoring, operating, responding,
surveillance, oversight, etc.) that they
perform to maintain the facility in a safe
operating or shutdown condition; (2) the
principal individuals by job title and a
summarized description of the human
actions to secure and protect the facility
(without providing sensitive
information); (3) the estimated total
population of individuals subject to the
FFD program and per shift by job
description; and (4) references to
supporting documentation. The purpose
of these descriptions is to enable an
NRC assessment of the licensee’s or
other entity’s analysis and the required
human actions to operate, monitor,
surveil, maintain, and secure the facility
within its design and licensing basis so
that if an operational or security-related
event were to occur, the facility would
respond as designed and licensed and
the calculated radiological dose
consequences would not exceed the
consequences described in
§ 53.860(a)(2). This is important because
facilities that implement § 26.604 are
expected to have very small staff sizes
and may be sited in geographically
remote locations, both of which could
challenge effective implementation of
the FFD program.
Proposed § 26.603(a)(2) would require
the applicant to state what FFD program
it plans to implement.
Proposed § 26.603(a)(3) would require
a discussion that informs the NRC of the
applicability of the applicant’s FFD
program to individuals who perform
safety- or security-significant activities.
This description should summarize any
key differences between the staff at the
site and any remote facility and the
categories of individuals in § 26.4. The
principal purpose of providing this
description would be to inform the NRC
of any substantial differences in the

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applicability of the FFD program to the
categories of individuals in § 26.4.
Proposed § 26.603(a)(4) would require
a description of the drug and alcohol
testing and fitness determination
process to be implemented through the
licensee’s or other entity’s procedures,
including the collection and testing
facilities to be used, biological
specimens to be collected, and sanctions
to be imposed upon a confirmed FFD
policy violation. This process includes
how individuals who test positive for a
drug or alcohol will be evaluated before
being afforded unescorted access to the
protected area to perform or direct those
duties or responsibilities making them
subject to the FFD program. The
principal purpose of describing this
return-to-duty process is to inform the
NRC of the behavioral observation
strategy (for those facilities that
implement § 26.604) and/or drug
screening and testing strategy.
Proposed § 26.603(a)(5) would require
a summary description of the
applicant’s planned PMRP. This
description must provide the
performance measures and thresholds
that the applicant intends to use.
Proposed § 26.603(b) would establish
when the FFD program must be
implemented and the longevity of the
FFD program. This proposal is
equivalent to the current § 26.3, which
states, in part, when licensees and other
entities must begin implementing their
FFD programs. Unlike the current part
26 regulations, proposed § 26.603(b)
would expressly state that an FFD
program would not be applicable during
decommissioning of a part 53 facility for
licensees and other entities specified in
§ 26.3(f). However, licensees of facilities
licensed to operate a reactor should be
aware that the physical protection
program under § 73.55, ‘‘Requirements
for physical protection of licensed
activities in nuclear power reactors
against radiological sabotage,’’ and
under proposed § 73.100 include a
requirement for the implementation of
an IMP, even during decommissioning.
Proposed § 26.603(b) would also
require the holder of an ML to
implement its FFD program no later
than the start of activities that assemble
a reactor. The holder of the ML should
establish in its procedures when reactor
assembly commences and what
constitutes assembly. For example, the
FFD program would not need to be
implemented for the receipt, storage,
inspection, and staging of components
and systems used to assemble (i.e., build
or fabricate) the reactor because this is
not a current requirement for LWR
facilities licensed under part 50 or 52.
Furthermore, the NRC currently does

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not require that an FFD program be
applied to the assembly or
manufacturing of components (or basic
components as defined in § 21.3), or
systems that were fabricated or
assembled outside the footprint of a
commercial power reactor, and this
regulatory position would also apply to
a manufacturing facility.
Proposed § 26.603(c) would require
the applicant, licensee, or other entity
seeking to implement an FFD program
under § 26.604 to perform a site-specific
analysis to determine whether the
facility and its operation satisfy the
criterion in § 53.860(a)(2). If the analysis
is performed and demonstrates that the
radiological consequences presented by
the facility and its operation satisfy the
criterion, then the licensee or other
entity could implement the FFD
program detailed in § 26.604. If the
analysis does not demonstrate that the
facility and its operation satisfy the
criterion, then the licensee or other
entity must implement the FFD program
described in either § 26.605 or subparts
A through I, N, and O of part 26.
Proposed § 26.603(c) would also
require licensees and other entities that
implement proposed § 26.604 to update
the technical analysis used to justify
compliance with the criterion in
§ 53.860(a)(2). This analysis would be
updated to reflect changes made to the
staffing, FFD programs, or offsite
support resources described in the
analysis to show that the facility and its
operation continue to satisfy the
criterion. This is important because
facility, operation, or staffing changes
outside FFD program implementation
(e.g., changes in the facility safety
analysis, physical protection strategies,
or the security plan, implementing
procedures, or contingency response
strategies) could adversely impact the
licensee’s or other entity’s documented
analysis demonstrating that the facility
and its operation satisfy the criterion if
event sequences require human action.
Proposed § 26.603(d) would require
the establishment of a PMRP. The
concept of a PMRP is not new. This
requirement would consolidate for part
53 the requirements in current §§ 26.41,
‘‘Audits and corrective actions’’; 26.415,
‘‘Audits’’; 26.717, ‘‘Fitness-for-duty
program performance data’’; and
26.183(c), which describes MRO
responsibilities. The proposal would
state that the licensee or other entity
must monitor the effectiveness of its
FFD program by comparing performance
data against performance measures and
thresholds. The development of
quantitative thresholds would be new,
but this is born from licensees and other
entities with facilities licensed under

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parts 50 or 52 already collecting,
reviewing, and reporting FFD
performance data. Additionally, the
benefit of quantitatively measuring FFD
program performance against
established thresholds benefits a
licensee’s and other entity’s
determination of whether they are
maintaining FFD program performance
in a manner that demonstrates
compliance with the performance
objectives in § 26.23.
The NRC is proposing the PMRP
because the subpart M of part 26
requirements would enable a high
degree of flexibility in FFD program
implementation (e.g., drug testing). A
licensee or other entity would not only
have options in the type of FFD program
they may implement under part 26, but
they would have options in the types of
biological specimens they may test for
drugs, where to collect the biological
specimens (e.g., at the NRC-licensed
facility or offsite at a local hospital or
clinic), and the use of collection and
assessment devices to screen
individuals for drugs and alcohol. These
FFD program flexibilities could cause
FFD programs under subpart M of part
26 to become very site-specific,
necessitating performance measures to
enable the licensee or other entity to
maintain the effectiveness of its FFD
program.
Fitness-for-duty program effectiveness
would be determined by comparing
actual performance against the
performance measures and thresholds.
The result of that comparison would
inform licensee or other entity decisions
whether to change FFD program
elements to address a performance
deficiency. Also, the thresholds would
have sufficient margin, based on
operating experience, before conditions
adverse to safety and security may occur
should an individual be identified as
impaired or not trustworthy and
reliable. The potential of a humanrelated failure causing a condition
adverse to safety and security is
dependent on the duties and
responsibilities of the individual and
the defense-in-depth designed to
prevent or mitigate an adverse
consequence. The PMRP would account
for this by requiring the review of FFD
performance data, in part, by work
category, C/V, and individuals
employed by the licensee who are not
a C/V as defined in § 26.5 (i.e., a
licensee employee).
Proposed § 26.603(d)(1) would require
the licensee or other entity to document
and maintain its PMRP. Proposed
§ 26.603(d)(1)(i) would require that the
performance measures be identified and
designed to monitor FFD program

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performance. Proposed
§ 26.603(d)(1)(i)(A) would require the
FFD program of a licensee or other
entity subject to the requirements of
§ 26.604 to include monitoring of the
BOP. The purpose of this monitoring is
to help ensure that individuals subject
to the FFD program are observing the
behaviors of others, are being observed
themselves, and are reporting FFD
concerns to licensee- or other entitydesignated individuals. The other
performance measures would include
occurrence of FFD policy violations
evaluated by licensee employee, C/V,
and labor category, and occurrence of
individuals with potentially
disqualifying information or who
possessed an FFD prohibited item.
Proposed § 26.603(d)(1)(i)(B) would
require the FFD program of a licensee or
other entity that is either subject to the
requirements of § 26.604 and has
implemented a drug testing program at
its discretion, or is subject to the
requirements of § 26.605, to include the
performance measures identified in
§ 26.603(d)(1)(i)(A) and those necessary
to monitor the effectiveness of the drug
and alcohol testing program. The drug
and alcohol measures would include the
monitoring of FFD performance data for
pre-access and random testing and
subversion attempts by the categories of
licensee employee, C/V, and labor
category.
Proposed § 26.603(d)(1)(ii) would
require the licensee or other entity to
establish thresholds for each
performance measure. Initial thresholds
must be based on FFD performance data
from comparable facilities subject to
part 26, the licensee’s or other entity’s
fleet-level program performance if
applicable, and industry FFD
performance data. This provision
introduces the requirement to ‘‘maintain
FFD program effectiveness.’’ This
terminology describes a performancebased regulatory strategy in which the
licensee or other entity must initially
establish a level of performance that is
representative of other facilities in the
licensee’s fleet of facilities subject to
part 26, if applicable, and the FFD
performance of comparable facilities
subject to part 26.
Proposed § 26.603(d)(1)(iii) would
require that the licensee or other entity
evaluate FFD data as it is received to
determine whether a threshold has been
exceeded. Historical FFD performance
data for the current LWR fleet indicates
that, for particular work categories and
employment types, few FFD policy
violations occur per year. Therefore, for
work categories that may be significant
to worker safety (e.g., radiation
protection technicians), physical

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protection (i.e., security personnel), or
safety (i.e., NRC-licensed operators and
individuals who perform or direct the
performance of activities that a riskinformed evaluation process has shown
to be significant to public health and
safety), a single FFD policy violation
could be a significant occurrence and
warrant corrective actions. Based on
licensee-submitted FFD-related reports
under §§ 26.417, 26.419, 26.717, and
26.719, licensees and other entities with
facilities licensed under parts 50 or 52
implement some form of corrective
action that is typically scaled to the
significance of the violation. These
corrective actions have included
counseling, follow-up drug and/or
alcohol testing, remedial training,
generic announcements to the
workforce, and reviews of recently
performed or directed work by the
individual suspected of being impaired.
Proposed § 26.603(d)(1)(iii) would
require that the PMRP include a year-toyear comparison of FFD performance
data to help provide assurance that an
adverse trend in FFD program
performance would be identified if
occurring. This proposed requirement
was developed from the annual FFD
performance data reporting
requirements in §§ 26.417(b)(2) and
26.717. In particular, the proposed yearto-year comparison of FFD performance
data is equivalent to § 26.717(c), which
requires, in part, licensees and other
entities to analyze their performance
data at least annually and take
appropriate actions to correct any
identified program weaknesses.
Proposed § 26.603(d)(1)(iv) would
require the licensee or other entity to
perform and document quantitative and
qualitative reviews. These reviews
would be performed in three program
areas: protections afforded to
individuals subject to the FFD program,
laboratory test results and MRO
performance, and change control. The
purpose of these reviews would be to
specifically target performance within
the three program areas to assess
whether the outcomes resulting from the
implementation of procedure
requirements are contributing to FFD
program effectiveness. The proposed
reviews would not require the
establishment of measures and
thresholds because the reviews are
expected to result in qualitative findings
regarding program effectiveness.
Qualitative findings and observations
could still result in the consideration of
corrective actions in the targeted
program areas.
Proposed § 26.603(d)(1)(iv)(A) would
require the licensee or other entity to
monitor whether its FFD program is

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affording appropriate protections to
individuals subject to the FFD program.
The review of these protections would
include, in part, assessing the licensee’s
or other entity’s protection of the
following: privacy during the specimen
collection process; specimen integrity,
custody, and control; information
gathered from FFD program
implementation; and due process during
appeals of FFD policy violations.
Proposed § 26.603(d)(1)(iv)(B) would
require, in part, a review of laboratory
test results and MRO performance.
Effective performance by the laboratory
(e.g., obtaining and communicating
accurate test results) and MRO (e.g.,
correct evaluation of the laboratory test
results based on § 26.185 or HHS
Guidelines) would result in three
significant outcomes: (1) protection of
the donor from an inaccurate FFD
policy violation determination; (2)
protection of the donor, other
individuals, and the facility from
potential harm should the donor be
impaired or not trustworthy and
reliable; and (3) a performance-based
assessment of both the laboratory and
MRO. This last outcome could facilitate
actions to improve laboratory
performance, MRO training under
§ 26.607(m), or both. Proposed
§ 26.603(d)(1)(iv)(B) would also require
a comparative analysis between the
POCTA screening result(s) and the
corresponding specimen test results
obtained from the HHS-certified
laboratory if the POCTA indicated a
positive, adulterated, substituted, or
invalid screening result or discrepant
biological marker, to assess the
effectiveness of the POCTA and to
inform MRO decisions under § 26.185 or
§ 26.607(m)(6). The results of this
biennial review could also inform the
conduct of laboratory audits.
Proposed § 26.603(d)(1)(iv)(C) would
require that the change control
requirement in proposed § 26.603(e) be
included in the biennial program review
to help ensure that changes
implemented over the life of the facility
do not result in a reduction in program
effectiveness even if a mitigating action
was implemented for the specific
change. This requirement was
developed from §§ 26.137(f) and
26.713(d). This part of the review would
require an assessment of all changes
since the last review and their potential
aggregated impact on FFD program
effectiveness. For example, if last year
the licensee elected to contract with a
different MRO and this year the licensee
implemented a new type of POCTA
device, each of those program changes
probably would not have resulted in a
recognizable reduction in FFD program

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effectiveness. But, if the drug testing
positivity rate (or FFD policy violations)
for C/Vs decreased markedly during a
future maintenance outage that required
many C/Vs, then the reduction could
indicate, for example, that the POCTA
device was not as effective as
determined by a forensic toxicologist
review under §§ 26.603(e) and 26.607(h)
or that the new MRO was improperly
crediting prescription medication for
laboratory-confirmed positive test
results.
Proposed § 26.603(d)(2) would state
when the licensee or other entity must
implement corrective actions. This
requirement would be equivalent to the
requirement in current § 26.415(b) and
was developed from requirements
contained in §§ 26.41(a) and (f),
26.127(e), 26.129(b)(1)(i), 26.137(f)(3)
through (5), 26.155(a)(6), 26.157(e),
26.159(b)(1)(i), and 26.203(e)(2).
Corrective actions must be implemented
to correct root causes, contributing
causes, or both. There is margin built
into the FFD performance thresholds
and qualitative factors (e.g., to account
for potential changes in drug and
alcohol testing performance data when
there is a large influx of C/Vs to perform
maintenance) that may influence a
licensee or other entity’s causal
determination for an occurrence. Thus,
generalized or qualitative corrective
actions may be implemented like
informing management and placing a
sufficiently descriptive summary of the
occurrence in a corrective action
program for future monitoring to assess
recurrence.
However, should the occurrence
challenge safety or security or
significantly exceed a performance
threshold even when considering
qualitative factors and margin, the
licensee or other entity should
implement more robust corrective
actions to resolve the cause. An example
of a challenge to safety or security
would be the situation when an NRClicensed operator or maintenance
professional had operated, surveilled, or
maintained safety-significant SSCs and
was determined to have been impaired
by behavioral observation or potentially
under the influence of a narcotic as
determined by an alcohol or drug test or
screening result. Immediate corrective
actions could include, but would not be
limited to, a licensee or other entity
assessment of the duties and
responsibilities recently performed by
the individual. Operating experience
within the LWR operating reactor
community demonstrates few FFD
policy violations per year per site have
been caused by individuals who
perform or direct the performance of

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safety or security-significant activities.
Therefore, any such violations of the
FFD policy in a particular work category
in one year could be a significant
performance deficiency. These
violations could be even more
significant at part 53 facilities that have
a very small workforce subject to part
26.
Proposed § 26.603(d)(3) would require
the licensee or other entity to biennially
assess and document its FFD
performance monitoring program; this
requirement was developed from
§ 26.41(b). This documented review
would demonstrate that the
performance measures and thresholds
are appropriate based on site- and
licensee’s fleet-level program
performance, if applicable, and industry
performance and adjusted to maintain
FFD program effectiveness. Also, as a
result of this effort, the licensee or other
entity would be in possession of lessons
learned from fleet-level performance, if
applicable, and industry performance
that could contribute to their own
performance assessment to maintain
program effectiveness.
Under proposed § 26.603(d)(3)(i), the
identified program weaknesses and
corrective actions resulting from the
biennial review would be required to be
summarized in the licensee’s or other
entity’s annual report to the NRC in
compliance with either § 26.417(b)(2) or
§ 26.717, as applicable. This information
would inform the NRC of FFD program
weaknesses to facilitate regulatory
oversight and enable the NRC to
aggregate industry data for use in a
licensee or other entity PMRP.
Proposed § 26.603(d)(3)(ii) would
establish when the biennial PMRP
review must be completed and when
corrective actions from the review must
be implemented. The NRC selected the
May 15th date of odd-numbered years to
help ensure that all FFD programs will
maintain their previously determined
performance measures and thresholds or
reset them based on FFD program
performance early in the year in which
the biennial review was conducted. This
would assist in obtaining quality FFD
performance data over two annual
reporting cycles and evaluating whether
previous corrective actions were
effective.
In proposed § 26.603(e), the NRC
proposes a change control requirement
for subpart M of part 26 FFD programs.
Requiring licensees and other entities to
demonstrate compliance with certain
requirements before implementing
changes to their FFD programs would be
necessary for two primary reasons. First,
proposed changes to a licensee’s or
other entity’s FFD program could affect

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the analysis performed by the licensee
or other entity under proposed
§ 26.603(c), which helps determine the
FFD program requirements that must be
implemented. If this analysis changes,
then the licensee’s or other entity’s FFD
program requirements might change.
Second, the requirements in subpart M
of part 26 are performance based.
Therefore, FFD program implementation
may change periodically in response to
societal changes in substance abuse or
from PMRP implementation. Change
control therefore relies on the licensee
or other entity maintaining its
procedures in a manner that details how
its FFD program is to be implemented
while incorporating changes, with
documentation that justifies the changes
to support the PMRP, audits, and NRC
inspection.
Proposed § 26.603(e)(1) would permit
the licensee or other entity to
implement changes to its FFD program
if it performs and retains an analysis
demonstrating that the change does not
reduce the effectiveness of the FFD
program or the change was necessitated
or justified by a change to part 26,
laboratory processes, or guidance issued
by the HHS or NRC. The proposed
change control requirement would
enable flexibility in program
implementation should the NRC or HHS
change its drug testing procedures (as
implemented by the licensee or other
entity through its procedures) in
response to changes in societal
substance abuse or drug testing
technologies.
The proposed change control
requirement was developed from the
change control requirements in
§ 50.54(p) and (q)—the change control
requirements for security and
emergency plans, respectively.
However, unlike these two
requirements, the NRC does not review
and approve a licensee’s or other
entity’s FFD program or its
implementing procedures, and the FFD
program is not licensing-basis
information as described in § 53.1300.
Proposed § 26.603(e)(2) would require
that if a change reduces FFD program
effectiveness, then the licensee must
implement a mitigating strategy so the
FFD program, as revised, will continue
to demonstrate compliance with the
performance objectives in § 26.23 and
not result in a reduction in program
effectiveness.
Proposed § 26.603(e)(3) would
prohibit, with one exception, the use of
the change control process to reduce the
minimum panel of drugs to be tested
and would reference the drugs listed in
proposed § 26.607(c)(1). Proposed
§ 26.607(c)(1) would reference current

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§ 26.31(d)(1), which states that, at a
minimum, licensees and other entities
shall test for marijuana metabolite,
cocaine metabolite, opioids (codeine,
morphine, 6-acetylmorphine,
hydrocodone, hydromorphone,
oxycodone, and oxymorphone),
amphetamines (amphetamine,
methamphetamine,
methylenedioxymethamphetamine, and
methylenedioxyamphetamine),
phencyclidine, and alcohol. The testing
of these drugs and drug metabolites,
except phencyclidine, and alcohol is
necessary for the FFD program to
remain effective. Also, there is no
proposed subpart M of part 26
requirement stating that this panel of
drugs and drug metabolites needs to
consist of only scheduled drugs.10 This
flexibility would account for the
situation where an impairing substance
becomes prevalent in society and a
licensee or other entity elects to add the
substance to their panel of substances to
be tested prior to it being scheduled by
the Drug Enforcement Administration.
The exception in proposed
§ 26.603(e)(3) would be that, should
HHS elect to remove phencyclidine
from the panel of drugs and drug
metabolites to be tested, a licensee or
other entity could make this change in
its FFD program without resulting in a
reduction in FFD program effectiveness.
This outcome would be justified based
on the very infrequent occurrence rate
of FFD policy violations due to
phencyclidine use since 2010. However,
if HHS proposes to remove a class of
drugs from the panel of drugs to be
tested that is listed in § 26.31(d)(1),
except for phencyclidine, then a
licensee or other entity may not make a
similar change to its panel of drugs to
be tested, because this change would be
a reduction in FFD program
effectiveness even with a mitigative
strategy implemented.
Changes in the HHS panel of drugs
and drug metabolites to be tested may
also shift from one metabolite to a
10 The Drug Enforcement Administration
classifies drugs, substances, and certain chemicals
used to make drugs into five (5) distinct categories,
depending upon the drug’s acceptable medical use
and the drug’s abuse or dependency potential.
These categories appear as Schedules I through V
of section 202 of the Controlled Substances Act (21
U.S.C. 812). Schedule I drugs have a high potential
for abuse, have no currently accepted medical uses
in treatment in the United States, and lack accepted
safety for use under medical supervision. At the
other end of the classification scheme, Schedule V
drugs have the least potential for abuse among the
five categories of drugs, have a currently accepted
medical use in treatment in the United States, and
abuse of the drug may lead to limited physical
dependence or psychological dependence. For more
information, see https://www.dea.gov/druginformation/drug-scheduling.

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different metabolite for the same drug
class (e.g., amphetamines, opioids) to be
tested. Should HHS issue such a change
to its panel, this would not be expected
to result in a reduction in FFD program
effectiveness because HHS would be
targeting a more prevalent or effective
metabolite in its drug testing program.
This situation could occur as HHS
gathers more operating experience from
Federal Government implementation of
its HHS Guidelines, or data generated by
drug testing laboratories and federally
mandated drug testing programs
required by Federal agencies such as the
NRC and U.S. Departments of
Transportation, Energy, and Defense.
Proposed § 26.603(e)(4) would require
that change control records be
maintained for a 5-year record retention
period based on the current NRC
practice to conduct triennial inspections
of licensees’ and other entities’ FFD
programs. This would afford the NRC an
opportunity to review the licensee’s or
other entity’s determination that FFD
program changes have not reduced the
effectiveness of their FFD program.
Licensees and other entities would also
be required to summarize each change
made under proposed § 26.603(e) in
their annual FFD performance reports
required by § 26.617(b)(2) or § 26.717, as
applicable.
Proposed § 26.604 would establish the
minimum set of FFD program
requirements for licensees and other
entities who have a documented
analysis that demonstrates that the
facility and its operation satisfy the
criterion in § 53.860(a)(2). For these
licensees, compliance with the
performance objectives in § 26.23 would
be ensured through the BOP; defense-indepth measures proposed in subpart M
of part 26 like the PMRP, change
control, and audits; and other
requirements, such as those for AA,
physical protection, and licensed
operators. The adequacy of these
measures in satisfying the performance
objectives is supported by operating
experience, which demonstrates margin
between an FFD-related occurrence and
a condition adverse to safety or security,
as illustrated by for-cause, post-event,
and random testing data. A facility that
satisfies the criterion in proposed
§ 53.860(a)(2) would present a smaller
potential radiological consequence than
a facility that does not satisfy the
criterion, so the requirements in
proposed § 26.604 are scaled to the
lower risk presented consistent with the
Commission’s Advanced Reactor Policy
Statement.
The disadvantages of implementing
the FFD program described in proposed
§ 26.604 would be few. Since drug and

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alcohol testing would not be required,
behavioral observation would be the
keystone requirement in this
performance-based framework to
provide that individuals are fit for duty,
trustworthy, and reliable, and can safely
and competently perform the duties and
responsibilities making them subject to
the FFD program. If not, the individuals
would be assessed in accordance with
the licensee’s or other entity’s
procedures similar in manner to that
required by subpart K of part 26, and the
proposed PMRP would require
corrective actions should a threshold be
exceeded.
If a licensee or other entity elects not
to perform the analysis in proposed
§ 26.603(c) to determine whether it
satisfies the criterion in proposed
§ 53.860(a)(2); performs the analysis and
finds that the facility and its operation
does not satisfy the criterion in
proposed § 26.603(c); or is a holder of an
ML, the licensee or other entity could
not implement the FFD program
described in § 26.604. Instead, the
licensee or other entity would
implement either the program described
in proposed § 26.605 or an FFD program
that demonstrates compliance with all
the requirements in current subparts A
through I, N, and O of part 26.
Proposed § 26.605 would establish
requirements in a graded manner
similar to the regulatory framework
established by the requirements in
subparts A through I, N, O, and K of part
26. This existing graded approach
consists of an FFD program for
construction of a commercial nuclear
plant and a more robust program that
must be implemented before reactor
operation. The former is the FFD
program in proposed § 26.605(a), and
the latter is proposed § 26.605(b). Like
that for an FFD program under § 26.604,
the FFD program under § 26.605 would
include FFD program elements similar
to those in subpart B of part 26, but the
proposed requirements are less
prescriptive, enabling more flexibility in
program implementation like that
offered in subpart K of part 26. For
example, the requirements in subpart B
of part 26 are explicit requirements for,
in part, the collection and analysis of
urine specimens. Subpart B of part 26
does not enable the use of oral fluid for
drug testing or screening, except under
very limited situations as described in
subpart E of part 26, or the use of hair
specimens, unlike proposed § 26.605.
Proposed § 26.605 would require drug
and alcohol testing based on either the
requirements in part 26 or the HHS
Guidelines. The principal benefit of the
proposed § 26.605 FFD program is that
it would provide a regulatory framework

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that is consistent with the radiological
consequences for a facility that does not
satisfy the criterion in proposed
§ 53.860(a)(2) while affording
flexibilities in the conduct of drug and
alcohol testing.
Proposed § 26.605(a) would apply to
licensees and other entities who
perform the § 26.603(c) analysis and
satisfy the criterion in § 53.860(a)(2) but
decide not to implement the FFD
program described in proposed § 26.604,
licensees and other entities who do not
perform the § 26.603(c) analysis, and
licensees and other entities who
perform the analysis but their analysis
does not demonstrate that their facility
and its operation satisfy the criterion in
§ 53.860(a)(2). These entities must
establish, implement, and maintain an
FFD program under § 26.605(a) either
during construction activities as defined
in § 26.5, or during activities performed
under an ML that allows the assembly,
testing, or both, of a manufactured
reactor. This FFD program implements
all the FFD program requirements in
§ 26.604 plus drug and alcohol testing.
The timing element of the proposed
applicability statement of § 26.605(a) is
equivalent to that for an LWR licensee
or other entity who is performing those
same activities at a facility licensed
under part 50 or 52 and helps provide
assurance that those individuals who
assemble, test, or perform construction
activities as defined in § 26.5 or direct
these activities are fit for duty and
trustworthy and reliable. This is
important because assembly and testing
a manufactured reactor and the
construction and testing of SSCs
required for facility operation require, in
part, adherence to procedures, possible
implementation of unique and precise
assembly techniques, and quality
assurance and controls. Additionally,
SSCs within a manufactured reactor
may not be accessible, testable, or
available for quality assurance and
verification after the reactor is
assembled. This requirement is also
proposed to address solo-assembly
activities that may cause latent failures
and passive SSCs located internal to a
reactor (for example, a fusible link
designed to melt at a particular
temperature to trigger an actuation
mechanism) that are relied upon for safe
operation but cannot be inspected or
tested for proper installation,
configuration, or operation after
installation. A § 26.605(a) FFD program
for these types of activities is equivalent
to the FFD program applicable to the
assembly of the reactor vessel internals
and testing of the SSCs internal to the
reactor at an LWR licensed under part
50 or 52.

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Proposed § 26.605(b) would apply to
the same licensees and other entities as
in proposed § 26.605(a) but before the
loading of fuel onsite into a reactor
vessel; before receiving a manufactured
reactor; or before individuals subject to
part 26 operate, test, perform
maintenance of, or direct the
maintenance or surveillance of securityrelated equipment or equipment that a
risk-informed evaluation process has
shown to be significant to public health
and safety. These entities must
establish, implement, and maintain an
FFD program that implements all the
requirements in § 26.605(a), except
proposed §§ 26.610, ‘‘Sanctions’’;
26.617, ‘‘Recordkeeping and reporting’’;
and 26.619, ‘‘Suitability and fitness
determinations’’; plus additional
requirements due to the increased
radiological consequences presented by
a part 53 commercial nuclear plant as
the licensee readies it for operation.
These additional requirements include
those in subparts C, D, H, and N of part
26, some of which would replace
§§ 26.610, 26.617, and 26.619.
Proposed § 26.605(b) would also
enable the licensee or other entity to
better integrate its facility into the LWR
fleet and Category I fuel cycle facilities
because subparts C, D, and H of part 26
would be required. These subparts
would be required, in part, because it is
expected that: (1) individuals will be
able to work at any part 50, 52, or 53
commercial nuclear plant and will
possess a nuclear safety culture and
desirable qualifications, skills,
expertise, or services; and (2) licensees
and other entities of facilities licensed
under parts 50, 52, and 70 may venture
to construct or operate a facility
licensed under part 53. Therefore, the
implementation of these subparts would
help ensure that all individuals subject
to part 26, except those individuals
subject to an FFD program under
§ 26.604, § 26.605(a), or subpart K of
part 26, would be subject to FFD
programs that provide reasonable
assurance that the individuals are fit for
duty, trustworthy, and reliable.
Proposed § 26.606, ‘‘Written policy
and procedures,’’ would require
licensees and other entities to
implement and maintain an FFD policy
and procedures for their FFD programs.
This section would establish
requirements equivalent to those in
current § 26.403, ‘‘Written policy and
procedures,’’ of subpart K. However, a
principal difference is that proposed
§ 26.606 is written to enable the use of
urine, oral fluid, and hair for drug
testing and screening.
Proposed § 26.606(a)(1) would require
each licensee and other entity to

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provide a written FFD policy statement
to individuals subject to the FFD
program before the individuals are
subjected to behavioral observation and
any FFD program drug and alcohol test.
This would be a protection measure
afforded to individuals subject to the
FFD program to help ensure that they
know what is expected of them before
being subject to the FFD program and
potential consequences should they
violate the FFD policy or procedures.
This requirement would also contribute
to safety and security because
understanding FFD program
responsibilities may enhance an
individual’s safety culture or the
individual may self-select out of the
licensee’s or other entity’s hiring
process.
Proposed § 26.606(a)(2) would require
that the FFD policy statement describe
the performance objectives in § 26.23,
which are the same FFD program
performance objectives required for
facilities licensed under parts 50, 52, or
70. Having a standard performance
outcome based on a licensee or other
entity satisfying the § 26.23 performance
objectives would enhance consistency
in FFD program implementation across
all entities subject to part 26. It would
also generate confidence that
individuals subject to part 26 will safely
and competently perform their duties
and responsibilities and use NRClicensed materials in a manner that will
protect the public health and safety and
common defense and security.
Proposed § 26.606(a)(3) would require
that the FFD policy statement describe
the minimum days off requirements in
§ 26.205(d)(3) or maximum average
work hours requirements in
§ 26.205(d)(7).
Proposed § 26.606(a)(4) would require
the FFD policy statement be written in
sufficient detail to provide affected
individuals with information on what is
expected of them and what
consequences may result from a lack of
adherence to the policy, including those
elements described in § 26.603(b), part
26-required sanctions, and required
medical/clinical treatment and followup testing for FFD policy violations.
This requirement is equivalent to
§ 26.403(a) of subpart K but includes an
additional description of what the
policy statement must include. For
example, the policy would describe the
NRC-required sanctions to help deter
substance abuse and required medical/
clinical treatment and follow-up testing
for FFD policy violations. This
provision would provide a protection
measure by helping the individual get
the assistance they need and help

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ensure that the individual refrains from
substance abuse.
Proposed § 26.606(a)(5) would require
that the FFD policy statement describes
the individual’s responsibilities to
report for work in a physiological and
psychological condition that enables the
safe and competent performance of
assigned duties and responsibilities and
inform a licensee- or other entitydesignated representative when the
individual determines that this cannot
be accomplished.
Proposed § 26.606(b) would require
licensees and other entities
implementing a FFD program in
accordance with subpart M of part 26 to
establish, implement, and maintain
written procedures for their FFD
programs. This requirement would be
equivalent to that in § 26.403(b) of
subpart K.
Proposed § 26.606(b)(1) would
establish requirements for a subpart M
of part 26 FFD program in which the
licensee or other entity implements a
drug and alcohol testing program. This
provision would be equivalent to the
requirements in current § 26.403(b)(1) of
subpart K, but § 26.606(b)(1)(i) through
(iv) proposes additional clarity and
specificity that licensees and other
entities must detail in their procedures
to address new testing methods in
subpart M of part 26 that are not
permitted under the current part 26
framework. Clarity and specificity in
procedural instructions would support
consistent program implementation,
which protects all individuals subject to
the program.
Proposed § 26.606(b)(1)(iv) would
require that if the licensee or other
entity elects to use the HHS Guidelines
for the conduct of drug testing, the FFD
program procedures must include the
name of the specific HHS Guideline and
revision being implemented by the
licensee or other entity and a
description of the specific sections in
the guideline that are being
implemented, including specimen
collections, drug testing, laboratory
procedures, and evaluation of test
results. This requirement would help
ensure the following: the validity and
accuracy of drug testing because the
specimens would be subject to
laboratory testing that has been certified
by the HHS; protection of worker rights
equivalent to the privacy, information,
and due process protections afforded to
Federal workers under the HHS
Guidelines because the HHS Guidelines
are used in the Federally mandated drug
testing programs; consistency in
program implementation because all
individuals subject to the FFD program
would be subject to the same collection,

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testing, and evaluation processes; and
FFD program effectiveness because the
effectiveness of the HHS Guidelines
have been verified by HHS’s National
Laboratory Certification Program
(NLCP). Detailed procedures would
enhance MRO and FFD program
personnel reviews of individual test
results because instructions would be
provided for, in part, the evaluation of
specific test results (e.g., positive,
negative, biological markers), the
conduct of additional testing for invalid
or dilute specimens, and the assessment
of subversion attempts (e.g., adulterated
or substituted). This would benefit FFD
program effectiveness and help prevent
misunderstanding of program
requirements and processes.
Proposed § 26.606(b)(2) would require
licensees and other entities to include in
their written procedures the immediate
and follow-up actions that would be
taken, and the procedures that would be
used, in certain situations specified in
proposed § 26.606(b)(2)(i) through (vi).
Proposed § 26.606(b)(2) would be
equivalent to the requirements in
current § 26.403(b)(2), which provides
the same requirement under an FFD
program for construction for part 50 or
52 licensees and other entities. This
would help ensure the effectiveness of
the FFD program and its consistent
implementation, because part 53
licensed facilities would be
implementing procedures to address the
same requirements and with individuals
who would understand what is
expected of them no matter what part 53
facility they were assigned.
The situation specified in proposed
§ 26.606(b)(2)(i) would arise when
individuals subject to the FFD program
have been involved in the use, sale, or
possession of illegal substances, illegal
drugs, or illicit substances. This
provision would be equivalent to
current § 26.403(b)(2)(i), except that the
phrase ‘‘illegal drugs’’ would be
replaced with ‘‘illegal substances, illegal
drugs, or illicit substances.’’ Illegal
substances would include legal
substances used in a manner
inconsistent with Federal or State law.
The situation specified in proposed
§ 26.606(b)(2)(ii) would arise when
individuals who are subject to the FFD
program are impaired by any substance
or the consumption of alcohol as
determined by behavioral observation or
a test that measures blood alcohol
concentration, as defined in § 26.5.
Except for a few differences, this
provision would be equivalent to
current § 26.403(b)(2)(ii) of subpart K.
The NRC would not include the phrases
‘‘to excess’’ and ‘‘accurately’’ in
proposed § 26.606(b)(2)(ii). Subpart M of

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part 26 is a performance-based
framework that focuses on impaired
human performance, and for alcohol,
impairment is determined by behavioral
observation or by blood alcohol
concentrations exceeding the limits in
§ 26.103, ‘‘Determining a confirmed
positive test result for alcohol,’’ using an
evidentiary breath testing (EBT) device
for alcohol (not whether an individual
drank ‘‘to excess’’). If impairment is
determined by an individual’s behavior,
it must be based on physiological
indications of alcohol impairment.
These indications are well established
in medical, clinical, and law
enforcement organizations, and could be
used by the licensee or other entity
through its procedures and training.11
The NRC would include the phrase
‘‘illegal substances, illegal drugs, and
illicit substances’’ in proposed
§ 26.606(b)(2)(ii) based on operating
experience and the terminology in
current § 26.23(b). There are far more
substances that may cause impairment
than just drugs, drug metabolites, and
alcohol. The phrase ‘‘before or while
constructing or directing construction of
safety- or security-related SSCs’’ in
current § 26.403(b)(2)(ii) would not be
included in proposed § 26.606(b)(2)(ii)
because proposed § 26.606 would apply
during construction, operation, and
decommissioning, if applicable. The
NRC would include the term
‘‘behavioral observation’’ in proposed
§ 26.606(b)(2)(ii) because impairment
can be visibly or audibly observed in an
individual, and individuals subject to
subpart M of part 26 would be trained
in behavioral observation under
proposed § 26.608.
The situation specified in proposed
§ 26.606(b)(2)(iii) would arise when
individuals who are subject to an FFD
program that includes drug and alcohol
testing attempt to subvert the testing
process by adulterating or diluting
specimens (in vivo or in vitro),
substituting specimens, or by any other
means. Except for one difference, this
provision would be equivalent to
current § 26.403(b)(2)(iii). The NRC
would include the phrase ‘‘if drug and
alcohol testing is conducted’’ to address
the licensee or other entity who
implements § 26.604, which does not
require drug and alcohol testing. The
purpose underlying this requirement
has increased in significance since
issuance of the 2008 part 26 final rule
11 By ‘‘well established’’ the NRC means that
there are Federal, State, and non-governmental
organizations with reputable and scientifically
based resources available for a licensee or other
entity to use in its procedures or training to inform
individuals of the physiological indications of
alcohol impairment or intoxication.

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because subversion attempts have
accounted for about one-third of all FFD
policy violations every year since 2016.
The situation specified in proposed
§ 26.606(b)(2)(iv) would arise when
individuals, who are subject to an FFD
program that includes drug and alcohol
testing, refuse to provide a specimen for
analysis or refuse to follow instructions
provided by FFD program personnel.
Except for two differences, this
provision would be equivalent to
current § 26.403(b)(2)(iv). As with
proposed § 26.606(b)(2)(iii), the NRC
would include the phrase, ‘‘if drug or
alcohol testing is conducted,’’ to
account for an FFD program
implemented under § 26.604. The NRC
would include the phrase ‘‘or follow the
instructions provided by FFD program
personnel’’ based on an existing
requirement in § 26.89(c) that the
collector must inform the donor that if
the donor refuses to cooperate in the
specimen collection process, then such
refusal will be considered a refusal to
test and sanctions for subverting the
testing process will be imposed.
The situation specified in proposed
§ 26.606(b)(2)(v) would arise when
individuals who are subject to an FFD
program had legal action taken relating
to drug or alcohol use. This requirement
would be equivalent to current
§ 26.403(b)(2)(v).
The situation specified in proposed
§ 26.606(b)(2)(vi) would be when
individuals subject to an FFD program
demonstrated character or actions
indicating that the individual cannot be
trusted or relied upon to perform those
duties and responsibilities or maintain
access to NRC-licensed facilities, SNM,
or sensitive information. This includes
character traits beyond those attributed
to drug or alcohol use. This proposal
would help ensure that the licensee or
other entity will implement an FFD
program designed to demonstrate
compliance with the § 26.23(c)
performance objective that FFD
programs must provide ‘‘reasonable
measures for the early detection of
individuals who are not fit to perform
the duties that require them to be
subject to the FFD program.’’ An
individual who is not trustworthy and
reliable is not fit to perform or direct the
performance of those duties and
responsibilities or be afforded those
types of access that make the individual
subject to an FFD program.
This proposed requirement also
would help to align the subpart M of
part 26 BOP with the BOP implemented
under § 73.56(f) and proposed § 73.120
and the purpose of the IMP as described
in § 73.55(b)(9) and proposed

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§ 73.100(b)(9).12 The demonstrated
character and actions of an individual
can indicate whether the individual can
be trusted and relied upon to safely and
competently perform assigned duties
and responsibilities or be afforded those
types of access making the individual
subject to the FFD program. This holds
true for any demonstrated adverse
character indication or action on- or
offsite.
The phrase ‘‘character or actions’’
would be used in proposed
§ 26.606(b)(2)(vi) to focus on observed
examples that indicate an individual
subject to subpart M of part 26 may not
be fit for duty or trustworthy and
reliable. Character traits include but are
not limited to personality, temperament,
honesty, carelessness, apathy,
psychosis, and commitment to safety
culture. Assessment of an individual’s
character should consider the potential
for changes in these traits when
compared to a previous baseline.
Actions would include a physical or
verbal demonstration of a character trait
that could call into question an
individual’s fitness, trustworthiness, or
reliability. For example, the individual
does something physically, verbally, or
in writing (e.g., falsifying records,
driving while impaired, or harming or
threatening to harm oneself, others, or
property) that compels another
individual to conclude that the observed
individual cannot be trusted or relied
upon. Unlike the background
investigation and reviews of ‘‘character
and reputation’’ in § 73.56(d)(6) and
(k)(1)(v) and proposed § 73.120, which
are principally retrospective reviews of
an individual and may be based on
third-party information (i.e.,
information from individuals not
subject to NRC requirements), the
‘‘character or action’’ focus of proposed
§ 26.606(b)(2)(vi) would be a present
observation of an individual subject to
the FFD program and performed by an
individual who is also subject to the
FFD program. Whether the information
would be received from an individual
subject to the FFD program or someone
who is not subject to the FFD program,
the licensee or other entity would need
to review this information (i.e.,
determine if the information and its
source are credible) to determine
whether the individual should maintain
authorization.
12 The IMP must monitor the initial and
continuing trustworthiness and reliability of
individuals granted or retaining unescorted AA to
a protected or vital area and implement defense-indepth methodologies to minimize the potential for
an insider to adversely affect, either directly or
indirectly, the licensee’s capability to protect
against radiological sabotage.

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Proposed § 26.606(b)(3) would require
licensees and other entities to address in
their procedures the process, including
the duties and responsibilities of FFD
program personnel, to be followed if an
individual’s behavior or condition raises
an FFD concern. This provision would
also require a process to be conducted
when credible information is received
by the licensee or other entity that the
individual is not fit for duty,
trustworthy, and reliable.
With a few exceptions, proposed
§ 26.606(b)(3) would be equivalent to
current § 26.403(b)(3). Instead of the
phrase ‘‘while constructing or directing
the construction of safety- or securityrelated SSCs’’ in current § 26.403(b)(3),
the NRC would use ‘‘on the NRClicensed facility’’ in proposed
§ 26.606(b)(3) because this provision
would apply during commercial nuclear
plant construction, operation, and
decommissioning, if applicable, in
addition to holders of an ML as
described in § 26.3(f). The requirement
that the roles and responsibilities of
FFD program personnel be described
was developed from current §§ 26.4(g)
and 26.31(b) and operating experience,
which has demonstrated that clear job
descriptions help ensure that
individuals know who is designated by
the licensee or other entity to make
decisions regarding FFD program
implementation and who can be
approached when physiological or
psychological help is needed. This is
principally a protection consideration
afforded to individuals subject to the
FFD program.
The proposed requirement would also
include two conditions not found in
current § 26.403(b) that would clarify
the initiation of the fitness
determination process should an
individual’s behavior or condition raise
an FFD concern. The phrase,
‘‘impairment from any cause that in any
way could adversely affect the
individual’s ability to safely and
competently perform the individual’s
duties,’’ would reflect the § 26.23(b)
performance objective. The condition,
‘‘the receipt of credible information
indicating that the individual cannot be
trusted or relied on to perform those
duties and responsibilities making the
individual subject to this part,’’ would
reflect the § 26.23(a) performance
objective. In either case, as required by
§ 26.23(c), the FFD program must
provide reasonable measures for the
early detection of individuals who are
not fit to perform the duties that require
them to be subject to the FFD program.
Proposed § 26.606(b)(4) would require
licensees and other entities to have
written procedures that address the

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operation and oversight of an onsite or
offsite collection facility. This
requirement would be equivalent to
current §§ 26.403(b) and 26.405(e) and
is developed from § 26.41(b), which
states that each licensee and other entity
who is subject to subpart B of part 26,
shall ensure that the entire FFD program
is audited, which is part of a licensee’s
or other entity’s oversight of the facility,
and § 26.87(a), which states that each
FFD program must have one or more
designated collection sites that have all
necessary personnel, materials,
equipment, facilities, and supervision to
collect specimens for drug testing and to
perform alcohol testing. Having
procedures for the operation and
oversight of the onsite or offsite
collection facility would enhance
consistency in program implementation,
protect individuals subject to testing,
and account for the flexibilities afforded
in the types of biological specimens
than may be collected under an FFD
program subject to subpart M of part 26.
Section 26.606(b)(4), when used with
the PMRP described in § 26.603(d) and
the proposed audit requirement in
§ 26.605(a), would help maintain FFD
program effectiveness and prevent
subversion attempts at facilities that
may not be under the direct day-to-day
oversight of FFD program personnel.
Proposed § 26.606(b)(5) would require
licensees and other entities to have
written procedures that address the
fatigue management requirements in
§ 26.202(b), ‘‘Procedures,’’ and either
§ 26.205(d)(3) or (d)(7).
Proposed § 26.606(b)(6) would require
licensees and other entities to have
written procedures that provide
measures to prevent subversion of drug
and alcohol tests conducted onsite and
offsite. This proposal was developed
from § 26.27(c)(1).
Proposed § 26.607, ‘‘Drug and alcohol
testing,’’ would establish drug and
alcohol testing requirements for
licensees and other entities
implementing proposed § 26.604, at
their discretion, and licensees and other
entities implementing proposed
§ 26.605. Except for a few differences,
proposed § 26.607 would be equivalent
to current § 26.405, which requires
licensees and other entities
implementing an FFD program under
subpart K of part 26 to have a drug and
alcohol testing program that
demonstrates compliance with the
requirements in § 26.405(b) through (g).
The differences are commensurate with
the risk consequences presented by a
part 53-licensed facility as compared to
a part 50 or 52 nuclear power plant.
These proposed requirements would
improve flexibility in the conduct of

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drug and alcohol testing while
maintaining protections afforded to
individuals subject to the FFD program.
Proposed § 26.607(a) would require
licensees and other entities to obtain a
split specimen for all drug tests using
oral fluid or urine for all test conditions
in § 26.607(b), (h) and (j). Neither
current subpart K nor current subparts
B or E of part 26 require a split
specimen. However, the majority of the
LWR fleet uses split specimens for drug
testing and commercially available drug
screening products use a split specimen
technique. Since publication of the 2008
part 26 final rule, the HHS has issued
guidelines for urine and oral fluid that
require split specimens, and the draft
proposed HHS Guidelines for hair
requires split specimens, as well.
The required use of a split specimen
process would protect the individual
because, upon a donor-alleged
discrepant or questionable test result,
the donor may provide permission to
test the split specimen (specimen B) in
an effort to refute the laboratory test
results for specimen A. The requirement
also would enable the MRO to direct
laboratory testing of specimen B if
specimen A were invalid; though the
NRC expects specimens becoming
invalid at the laboratory to be a rare
occurrence as testing would be
conducted in HHS-certified laboratories
with trained collectors. In the event that
a specimen is determined to be invalid,
then the occurrence would likely
warrant further investigation by the
MRO and laboratory to identify the
cause. This protocol would be
equivalent to the special analysis testing
in current § 26.163(a)(2) for dilute
specimens in that additional laboratory
analysis is performed because of a
questionable test result.
If a split specimen is tested by an
HHS-certified laboratory, then the test
result from specimen B must be used as
part of the determination for an FFD
policy violation as required by
§ 26.185(n), ‘‘Evaluating results from a
second laboratory.’’ However, this is not
to say that the test results from
specimen A should be discarded. Since
the HHS-certified laboratory should
report all test results from all specimens
tested to the MRO, like the information
described in § 26.169, ‘‘Reporting
results,’’ test result differences between
specimens A and B can be used to
inform the MRO as to what should be
reported to the licensee or other entity
to either facilitate medical or clinical
assistance for the individual, inform an
FFD-policy violation determination, or
both.
The proposed § 26.607(a) requirement
would also state that if the licensee or

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other entity elects to use a POCTA
device for screening during random
testing or portal area monitoring (e.g.,
pre-access screening), a split specimen
would not need to be taken. The reason
for this exception would be that the
requirements in § 26.607(h)(4) establish
the process to be implemented when a
screening test indicates a presumptive
positive, adulterant, or a discrepant
biological marker, if applicable. This
process includes collecting and testing a
specimen for analysis at an HHScertified laboratory.
Proposed § 26.607(b) would require
the licensee or other entity to subject
individuals identified in § 26.202 to
drug and alcohol testing under the five
conditions listed in § 26.607(b)(1)
through (5). Proposed § 26.607(b) would
be equivalent to current § 26.405(c).
Proposed § 26.607(b)(1) would require
pre-access testing similar to current
§ 26.405(c)(1), which requires testing
before assignment to construct or direct
the construction of safety- or securityrelated SSCs. Unlike current
§ 26.405(c)(1), the proposed requirement
would not include the phrase,
‘‘construct or direct the construction of
safety- or security-related SSCs,’’
because, for licensees or other entities
under part 53, the pre-access test
condition applies to construction,
operation, and decommissioning, if
applicable, to help inform a licensee’s or
other entity’s authorization
determination. The proposal also would
use ‘‘pre-access’’ instead of ‘‘preassignment,’’ which is used in current
§ 26.405(c)(1).
A pre-access test would require the
collection of an oral fluid or a urine
specimen no more than 14 days before
the individual is granted unescorted
access. Although this change has roots
in the 2008 part 26 final rule, which
reduced the period within which preaccess testing must be performed from
60 days to 30 days or less, the 14-day
proposal is based on three lessons
learned from operating experience.
First, the 14-day period would be a
large enough window of time to collect
the specimen and evaluate test results
because licensees or other entities
typically receive laboratory test results
within 5 business days of laboratory
receipt of the biological specimen. At
the same time, the 14-day period would
be small enough to help ensure that the
test results are representative of the
individual’s forensic toxicology before
being granted authorization.
Second, the 14-day window would
enable the licensee or other entity to
conduct an unannounced pre-access
drug and alcohol screening using a hair
specimen or a POCTA. This would help

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prevent an individual from attempting
to subvert the drug and alcohol test by
temporarily abstaining from drug or
alcohol abuse or adulterating or
substituting their specimen to obtain a
non-positive test result.
Third, the NRC does not expect
licensees and other entities licensed
under part 53 to have the large and
periodic influxes of individuals (either
licensee employees or C/Vs) that LWRs
have to support facility operation,
maintenance, engineering design
changes, or nuclear refueling. Therefore,
these licensees or other entities would
not be periodically challenged to in-take
a large workforce within the proposed
14-day pre-access testing window.
Proposed § 26.607(b)(2) would require
the licensee or other entity to conduct
random drug and alcohol testing of all
individuals subject to the FFD program.
With one exception, this proposed
requirement would be equivalent to
current § 26.405(b). Section 26.405(b)
gives licensees and other entities that
implement an FFD program subject to
subpart K of part 26 the option to
impose random drug and alcohol
testing. Proposed § 26.607(b)(2) would
not offer that option because subpart M
of part 26, unlike subpart K, would not
allow a licensee or other entity to
implement a fitness monitoring program
under current § 26.406 instead of a
random testing program. The principal
reasons for not allowing this flexibility
would be that no licensee or other entity
has ever implemented a fitness
monitoring program (i.e., there is no
operating or regulatory experience on
which to judge the effectiveness of a
fitness monitoring program) and the
proposed subpart M framework already
uses behavioral observation to help
ensure FFD program effectiveness.
Supplementing the proposed § 26.609
BOP with an additional observation
technique (i.e., the fitness monitoring
program) would not result in a level of
deterrence or detection equivalent to
that which would be obtained through
behavioral observation and random drug
and alcohol testing.
Proposed § 26.607(b)(2)(i) through (v)
would provide specific requirements for
the conduct of a random testing
program. These paragraphs would be
equivalent to § 26.405(b)(1) through (4),
although with a few differences. The
similar provisions would be proposed in
§ 26.607(b)(2)(i), (b)(2)(iii), and
(b)(2)(iv).
The differing provisions would
include proposed § 26.607(b)(2)(ii),
which would refer to an ‘‘FFD program
procedure’’ instead of the reference to
an ‘‘FFD program policy’’ in
§ 26.405(b)(2) because procedures

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contain the instructions that implement
FFD program requirements, but the FFD
policy need not contain specific
instructions. Section 26.607(b)(2)(ii)
would also require individuals who are
selected for random testing to report to
the onsite collection site, as opposed to
the collection site in § 26.405(b)(2)
because alcohol metabolism necessitates
a relatively timely alcohol test. This
change is also proposed because the
NRC expects that part 53 licensees and
other entities may use a combination of
onsite (for random, for-cause, and postevent testing) and offsite (for pre-access,
post-event, and follow-up testing)
collection facilities for drug and alcohol
testing and may have to afford
reasonable accommodation to certain
individuals, which would add
complexity in the licensee’s or other
entity’s procedurally determined time
period in which an individual must
report to the collection facility.
Another difference from § 26.405(b)
would be proposed § 26.607(b)(2)(v),
which would establish the random
testing rate for the population of
individuals subject to testing. Subpart K
of part 26 does not establish a random
testing rate. The proposed requirement
would be equivalent to current
§ 26.31(d)(2)(vii), which requires that
the sampling process used to select
individuals for random testing provides
that the number of random tests
performed annually is equal to at least
50 percent of the population that is
subject to the FFD program. The NRC
would revise that slightly for proposed
§ 26.607(b)(2)(v) to require a 50 percent
random testing rate for the licensee
employee population and a 50 percent
random testing rate for the C/V
population. The NRC proposes this
change for two reasons.
First, although operating experience
has demonstrated that § 26.31(d)(2)(vii)
helps provide reasonable assurance that
individuals are fit for duty and
trustworthy and reliable through the
detection and deterrence of substance
abuse, this same operating experience
demonstrates that, on many occasions,
the C/V population has been tested at a
rate lower than 50 percent, even though
this population results in the majority of
all FFD policy violations. This bias
occurs because C/Vs are available for
testing only during short periods of time
or periodically throughout the year,
whereas licensee employees are
essentially always available for a test.
A second reason why the NRC is
proposing a different 50 percent random
testing protocol than in the current part
26 requirements is that the flexibilities
afforded to part 53 licensees or other
entities in subpart M of part 26 are not

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afforded to licensees or other entities
that must implement an FFD program
under subparts A through I, N, and O of
part 26. These flexibilities include
enabling the use of a POCTA device to
screen individuals during the random
testing process and the use of offsite
collection facilities for pre-access
testing. The potential reduction in FFD
program effectiveness caused by
licensee or other entity implementation
of these options would be offset by
subpart M requirements that mitigate
possible challenges to the FFD program,
such as the 50 percent random testing
rate for the licensee employee
population and 50 percent random
testing rate for the C/V population.
Proposed § 26.607(b)(3) would require
for-cause testing equivalent to that used
in current FFD programs implementing
§ 26.405(c)(2). The NRC would require
for-cause testing, like random testing, to
be conducted onsite to ensure that the
test is conducted as soon as reasonably
practicable. This is an important
consideration when for-cause testing for
alcohol or using oral fluid for drug
screening or testing because human
metabolism continually lowers the
concentrations of the drugs, drug
metabolites, and alcohol perhaps to
concentrations lower than the initial or
confirmatory testing cutoffs.
Additionally, for facilities that are sited
in geographically remote locations, an
offsite collection facility might be too far
away or not readily accessible.
Proposed § 26.607(b)(4) would require
post-event testing in a manner
equivalent to current § 26.405(c)(3) with
a few adjustments. For part 53 licensees
or other entities, the NRC proposes postevent testing under two conditions:
events involving human errors that may
have caused or contributed to the events
(proposed § 26.607(b)(4)(i)), and events
not involving human error that result in
adverse health consequences or damage
to any safety- or security-related SSC
(proposed § 26.607(b)(4)(ii)). The word
‘‘significant’’ would not be used in
§ 26.607(b)(4)(ii)(A) to describe the
‘‘illness or personal injury’’ as used in
§ 26.405(c)(3)(i) because
§ 26.607(b)(4)(ii)(A) would describe
which illnesses or injuries are covered.
Proposed § 26.607(b)(4)(ii)(B), unlike
§ 26.405(c)(3)(ii), would not use the
word ‘‘significant’’ to describe the
damage to safety- or security-related
SSCs because any damage to safety- or
security-related SSCs would require
testing within four hours of the event
unless immediate medical intervention
precludes the conduct of the test on the
individual(s) who caused or contributed
to the event. Proposed
§ 26.607(b)(4)(ii)(B) also would not use

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the word ‘‘construction’’ as in
§ 26.405(c)(3)(ii) because § 26.607(b)(4)
would apply to construction, operation,
and decommissioning, if applicable.
Proposed § 26.607(b)(4)(i) would
require the licensee or other entity to
define in its procedures the terms
‘‘human error’’ and ‘‘event.’’ These
terms may take on various meanings
and they are not defined in the current
or proposed rule, so the licensee or
other entity would be required to
describe or define these terms to help
ensure consistent implementation of
subpart M of part 26 and that the postevent test condition would be
consistently applied to all individuals
subject to the FFD program. The
§ 26.405(c)(3)(i) requirement that ‘‘the
event is recordable under the
Department of Labor standards
contained in 29 CFR 1904.7, and
subsequent amendments thereto,’’
would not be carried over to proposed
§ 26.607(b)(4). Instead, the NRC
proposes to prescribe the post-event test
conditions in § 26.607(b)(4), in part so
they would not change unless the NRC
amends the requirement.
Proposed § 26.607(b)(5) would require
follow-up testing. This requirement
would be equivalent to current
§ 26.405(c)(4), although the proposed
§ 26.607(b)(5) would further describe
follow-up testing. The NRC proposes to
describe follow-up testing as part of a
series of tests for drugs, alcohol, or both,
which are performed after an individual
subject to part 26 has violated the FFD
policy on substance use or abuse, or the
sale, use, or possession of illegal drugs.
Follow-up testing would be used to
verify an individual’s continued
abstinence from substance abuse. The
NRC would not include a reference to a
follow-up plan as in § 26.405(c)(4)
because the intent of a follow-up plan
is to conduct a series of drug tests,
alcohol tests, or both, to verify
continuing abstinence from substance
abuse. Nevertheless, individuals who
violate an FFD policy on substance use
or abuse, or the sale, use, or possession
of illegal drugs, should have a follow-up
plan that includes a definition of
‘‘abstinence’’ from the medical
professional prescribing the plan.
Proposed § 26.607(c) would provide
additional testing requirements. This
proposed requirement would be
equivalent to § 26.405(d) and would
require implementation of select
requirements from current subpart E of
part 26. The proposed requirements
would govern directly observed
collections, shy bladder situations,
special analysis testing, and alcohol
testing. These requirements would be
necessary to maintain FFD program

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effectiveness equivalent to that
currently implemented by the LWR
fleet.
Proposed § 26.607(c)(1) would require
validity testing and establish the
minimum panel of drugs and drug
metabolites to be tested. This panel
would be the same as those in
§§ 26.31(d)(1) and 26.405(d) because,
based on operating experience from
LWR FFD program implementation, this
panel has been determined to contribute
to a licensee or other entity satisfying
the FFD performance objectives in
§ 26.23(a) through (d).
Proposed § 26.607(c)(1) would differ
from § 26.405(d) because it would
require testing of oral fluid and urine
specimens for validity, including at
least one biological marker (developed
from an HHS Guidelines provision) and
one adulterant (equivalent to current
validity testing for urine specimens in
part 26). Section 26.405(d) requires that
urine specimens collected for drug
testing be subject to validity testing. The
addition of oral fluid validity testing is
important because, just as there are
publicly available kits to subvert a urine
drug test, kits that may be used to
subvert a drug test that uses oral fluid
as a biological specimen are also readily
available.
Proposed § 26.607(c)(2) would
include requirements that already exist
in the part 26 framework that provide
protections for individuals subject to the
FFD program and contribute to testing
effectiveness when collecting and
assessing a urine specimen. Specifically,
current § 26.115, ‘‘Collecting a urine
specimen under direct observation,’’
describes the exclusive grounds for
performing a directly observed
collection and the process to be
followed to protect the privacy of the
individual. Section 26.119,
‘‘Determining ‘shy’ bladder,’’ establishes
the process to be followed when a donor
is not able to produce a sufficient
amount of urine for testing, and
§ 26.163(a)(2) requires special analysis
testing when a specimen is dilute to
help prevent a subversion attempt.
Proposed § 26.607(c)(3) would require
implementation of all the current
alcohol testing requirements in § 26.91,
‘‘Acceptable devices for conducting
initial and confirmatory tests for alcohol
and methods of use,’’ through § 26.103,
‘‘Determining a confirmed positive test
result for alcohol.’’ Using the same
alcohol testing framework for parts 50,
52, 70, and 53 licensees and other
entities would provide for regulatory
consistency, protections for individuals
subject to the FFD program (e.g., the
quality controls and verification applied
to the EBT device), and FFD program

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effectiveness (e.g., accuracy of test
results). For alcohol testing, unlike drug
testing, there is a preponderance of
evidence that correlates blood alcohol
concentrations to impairment and
intoxication. Furthermore, FFD
performance data has demonstrated that
the time-dependent alcohol cutoffs in
§ 26.103 have increased the detection of
individuals who are under the influence
of alcohol. For these reasons, the current
alcohol requirements in part 26 are
proposed for FFD programs under
subpart M.
Proposed § 26.607(c)(4) would
establish additional testing
requirements. This proposal would be
equivalent to current § 26.405(f) for
facilities licensed under part 53 for the
conduct of drug testing. Unlike
§ 26.405(f), proposed § 26.607(c)(4)
would not reference validity screening
and initial drug and validity tests at
licensee testing facilities as this would
be required in proposed § 26.607(c)(1).
Another minor difference between
§ 26.405(f) and proposed § 26.607(c)(4)
would reflect the requirement in subpart
M of part 26 to use an HHS-certified
laboratory for all biological specimens
collected and not just for urine
specimens.
Consistent with § 26.405(f), proposed
§ 26.607(c)(4) would require the use of
an HHS-certified laboratory for all test
conditions listed in § 26.607(b), MROdirected tests, and the testing of a split
specimen. Further, HHS-certified
laboratory test results using urine or oral
fluid would be required for the issuance
of an FFD policy violation and part 26required sanction.
All drug testing would need to be
performed at an HHS-certified
laboratory to help ensure FFD program
effectiveness and to protect the donor
from a false positive test result and an
unwarranted FFD policy violation. The
donor would be protected because
laboratory procedures for specimen
accessioning, testing, custody and
control, and evaluation of test results
and the training and qualification of
laboratory personnel are evaluated by
HHS as part of the NLCP. This provides
assurance that the drug testing results
are accurate and attributed to the donor.
Urine, oral fluid, and hair specimens
may also be screened and tested for
drugs and alcohol as described in
§ 26.607. Drug and alcohol screening
results obtained from urine and oral
fluid specimens collected and analyzed
using a POCTA device and screening
results obtained from a hair specimen or
a portal monitor may only be used as
potentially disqualifying information for
a licensee’s or other entity’s
authorization determination (i.e., used

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to assess the fitness, trustworthiness,
and reliability of the individual). These
screening results may not be used for
the administration of an FFD policy
violation and sanction, except as
proposed §§ 26.607(i)(3) and 26.610 for
subversions, as defined in § 26.5, of the
drug and alcohol screening process.
There are three phrases or
requirements in § 26.405(f) that the NRC
does not propose to use in
§ 26.607(c)(4). The first is the phrase,
‘‘consistent with its standards and
procedures for certification,’’ regarding
the operation of an HHS-certified
laboratory, because the laboratory
would not be HHS-certified if it were
not following ‘‘its standards and
procedures for certification.’’ The
second is the requirement that urine
specimens that yield positive,
adulterated, substituted, or invalid
initial validity or drug test results must
be subject to confirmatory testing by the
HHS-certified laboratory, except for
invalid specimens that cannot be tested.
This requirement would not be used
because, under subpart M of part 26,
licensees or other entities would be
required to use an HHS-certified
laboratory. For a laboratory to be HHScertified, it must follow the HHS
Guidelines and include procedures that
describe when a specimen cannot be
tested. Lastly, the § 26.405(f)
requirement that other specimens that
yield positive initial drug test results
must be subject to confirmatory testing
by a laboratory that demonstrates
compliance with stringent quality
control requirements that are
comparable to those required for
certification by the HHS, would not be
used because subpart M of part 26
would require the use of an HHScertified laboratory.
Proposed § 26.607(c)(4) would require
the licensee or other entity to contract
with a primary and backup HHScertified laboratory. This provision
would help ensure that specimens are
processed and tested to maintain FFD
program effectiveness should the
primary laboratory be unable to perform
specimen testing. This would help
maintain protections afforded to
individuals subject to the FFD program
(e.g., should the donor or MRO request
testing of the split specimen, a different
laboratory could be used). This
requirement also would state that the
primary and backup laboratories must
have a different certifying scientist.
Having a back-up HHS-certified
laboratory and a different certifying
scientist would benefit the program and
donor because the drug testing
instruments, technicians, and certifying
scientist would be independent of the

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primary laboratory testing and review
process. The back-up HHS-certified
laboratory may be of the same corporate
entity as the primary laboratory.
Proposed § 26.607(c)(4) would also
state that the laboratory would be
subject to inspection or audit by the
licensee or other entity and that records
and documents must be provided and/
or able to be photocopied and removed
from the premises to support the
inspection or audit. This requirement
would be equivalent to current
§ 26.41(d) except that laboratories
would not be able to limit the use and
dissemination of documents copied or
taken from the laboratory by a licensee
or other entity. This is necessary to
ensure the continuing effectiveness of
FFD programs, because NLCP findings
and audit results could adversely
impact FFD program effectiveness.
Pertinent information includes and
should not be limited to NLCPidentified weaknesses (e.g., custody and
control, accessioning, instrumentation,
procedures, training, supervision,
review of test results, and resolution of
previously identified corrective actions)
that may impact the effectiveness of
FFD programs.
Proposed § 26.607(d) would help
protect the donor from mistakes made
during the drug and alcohol testing
processes and help ensure FFD program
effectiveness. The rule would require
the licensee or other entity to protect the
individual’s privacy and the integrity of
the specimen and to implement quality
controls to ensure that test results are
valid and attributable to the correct
individual. This requirement would be
equivalent to the first sentence of
current § 26.405(e), except that the word
‘‘stringent’’ was removed from the
phrase ‘‘stringent quality controls,’’
because the word ‘‘stringent’’ is not
defined.
Proposed § 26.607(e) would describe
the requirements for licensees and other
entities that use offsite collection
facilities. Consistent with current
§ 26.405(e), a licensee or other entity
would be able to conduct specimen
collections and alcohol testing at a local
hospital or other facility. Unlike
§ 26.405(e), proposed § 26.607(e) would
not restrict licensees and other entities
to use hospitals and other facilities that
meet the requirements in 49 CFR part
40, ‘‘Procedures for Transportation
Workplace Drug and Alcohol Testing
Programs,’’ because subpart M of part 26
is intended to provide flexibilities
beyond those in the current part 26
framework. Licensees and other entities
may use these Department of
Transportation requirements to inform
their procedures under § 26.606(b)(1) as

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long as the procedures do not conflict
with the requirements in part 26 or the
HHS Guidelines.
Proposed § 26.607(e) would also
require licensees and other entities to
audit offsite collection facilities before
their use and biennially to confirm that
the facility procedures are comparable
to those described in subpart E of part
26 or the HHS Guidelines for urine and
oral fluid. This proposed requirement is
based on current § 26.41(a) and (b). The
§ 26.607(e) audit requirement would be
a program effectiveness consideration
because offsite collection facilities may
not require vigilance of their collectors
(e.g., identification of subversion
attempts), diligence in the protection of
worker rights (e.g., privacy and
specimen custody and control), or
procedural compliance.
The offsite facility used by a licensee
or other entity under proposed
§ 26.607(e) would have to be licensed to
conduct specimen collections and
perform alcohol testing, and be audited,
by the State or a State-designated entity.
This requirement would help provide
assurance of adequate collection facility
performance and may help reduce the
burden on the licensee or other entity
and the collection facility. Crediting a
State audit (or State licensure, oversight,
or regulation) is established in
§§ 26.4(i)(4) and (j), 26.91(e)(5),
26.153(f)(1), and 26.183(a).
Proposed § 26.607(f) would provide
the requirements for initial drug testing.
This provision would be equivalent to
§ 26.405(f) except to account for the
alternative biological specimens that
may be tested under subpart M of part
26. For the testing of all biological
specimens, the licensee or other entity
under part 53 would be required to use
a device that employs an immunoassay
screening technique, or an alternative
technology that the licensee or other
entity has incorporated into its FFD
program through the § 26.603(e) change
control process, that demonstrates
compliance with the requirements of the
U.S. Food and Drug Administration
(FDA) for commercial distribution.
Examples of alternative technologies
include liquid or gas chromatography
and mass spectrometry. Licensees and
other entities would use the § 26.603(e)
change control process to evaluate and
document a change to their collection
and analysis procedures to enable the
use of a better or perhaps more costeffective collection and/or testing
technology. Another difference from
§ 26.405(f) would be changing the word
‘‘urine’’ in § 26.405(f) to ‘‘biological
specimens’’ in § 26.607(f). Lastly,
proposed § 26.607(f) would include the
phrase ‘‘discrepant biological marker’’

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as a drug screening result that must be
analyzed by an HHS-certified laboratory
and evaluated by the MRO to help
inform the MRO’s determination of a
subversion attempt.
Proposed § 26.607(g) would enable a
part 53 licensee to use oral fluid as a
biological specimen for testing. This
requirement would be equivalent to
§ 26.31(d)(5), which enables the MRO to
conduct drug and alcohol testing using
alternative methods, and § 26.405,
which does not preclude the use of oral
fluid specimens for FFD programs that
implement subpart K of part 26
requirements. In order to provide
assurance that drug testing is effective
and protects the worker, § 26.607(g)
would require that the licensee’s or
other entity’s procedures incorporate
the HHS Guidelines or the requirements
in part 26 for the conduct of urine or
oral fluid testing.
The proposed § 26.607(g) requires that
the oral fluid collection device must
have received premarket approval from
the FDA and must not expire before
laboratory testing. Also, the drugs, drug
metabolites, initial and confirmatory
testing cutoffs, and biological markers, if
applicable, must be those established by
HHS for oral fluid drug testing and the
alcohol cutoffs in part 26. If they are not
established by HHS or the NRC for the
paneled drugs and drug metabolites,
then they would be determined and
documented by a forensic toxicologist
review. This forensic toxicologist review
would help ensure that the device
accurately tests for the drug, drug
metabolite, biological markers,
adulterants, and/or alcohol and that the
results from the device are comparable
to those established in the HHS
Guidelines for oral fluid testing.
Proposed § 26.607(h)(1) and (2) would
enable the use of a POCTA device
during the random and pre-access
testing processes. These requirements
are adopted from § 26.97, ‘‘Collecting
oral fluid specimens for alcohol and
drug testing,’’ and § 26.405(f), which
does not preclude the use of oral fluid
testing. To use a POCTA device for
urine, oral fluid, or other biological
indicators (breath, sweat, etc.), a
forensic toxicology review would be
required to ensure that the device is
forensically effective. If the POCTA
device is forensically effective, then the
donor would be reasonably protected
from a false positive test result, the
licensee or other entity would be
reasonably protected from false negative
test results, and the FFD program would
remain effective. For a POCTA device to
be forensically effective, the forensic
toxicologist would need to document an
evaluation that the performance of the

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POCTA device must be comparable to
the requirements in § 26.161(b) for a
urine specimen or the procedures in the
HHS Guidelines for urine or oral fluid,
as implemented by the licensee or other
entity through its procedures.
The use of POCTA for oral fluid and
urine specimens for the pre-access and
random testing processes would be
acceptable because individuals in the
pre-access process would be subject to
an oral fluid or urine specimen
collection and possible drug screening
using a hair specimen, which are both
required to be sent to an HHS-certified
laboratory. For random testing, the
individual would have also been
granted authorization under the AA and
FFD requirements and have been subject
to behavioral observation and physical
protection screening (e.g., verification of
identify, and screening for explosives
and contraband).
Proposed § 26.607(h)(3) would require
that procedures be developed that
ensure the effectiveness of the POCTA
collection process, assessment of the
screening results, and prevention of
subversion attempts. This requirement
would be equivalent to current
§ 26.403(b)(1) and would help ensure
protections afforded to individuals
subject to the FFD program and program
effectiveness. The subpart M of part 26
framework enables the use of POCTA
for random screening of individuals for
any part 53 facility, so the licensee or
other entity should exercise due
diligence and implement risk
management strategies to ensure the
efficacy of random screening and its
contribution to an effective FFD
program.
Proposed § 26.607(h)(4) would
provide that an individual donor who
screens positive (or whose specimen is
invalid or indicates a discrepant
biological marker or adulterant) is
removed from all duties and
responsibilities making the donor
subject to subpart M of part 26. Under
proposed § 26.607(h)(4)(i), the donor
then would be immediately subject to a
drug and alcohol test that provides
quantified confirmatory test results from
which an FFD policy violation may be
issued. Similar to other requirements for
specimen collections, except for
biological specimens analyzed by a
passive detection system, the licensee or
other entity would be required to
implement procedures that ensure that
all specimens collected are uniquely
assigned to the donor (i.e., procedures
that provide for custody and control of
the specimen). If the individual shows
signs of impairment during the POCTA
process, proposed § 26.607(h)(4)(ii)
would require the temporary removal of

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the individual’s authorization until the
MRO reviews the laboratory test
result(s), and interviews the individual,
and a determination of fitness finds that
authorization may be restored. Section
26.607(h)(4) is equivalent to § 26.77(b)
and was informed by the requirements
in §§ 26.419, 26.75(c) and (d), and
26.185(c).
Proposed § 26.607(i) would enable the
collection of hair specimens for drug
testing to supplement pre-access testing
that uses urine or oral fluid specimens.
Hair testing would be a new feature in
the part 26 framework. The NRC
proposes to permit the use of hair
testing for only Schedule I or II drugs or
their metabolites to inform a licensee’s
or other entity’s determination whether
the individual is trustworthy and
reliable. For example, if an individual
stated no prior use of illegal drugs or
potentially addictive habits, a hair
screening test could be performed
during the pre-access process to
ascertain the validity of the individual’s
statement. However, if the HHS-certified
laboratory communicates a laboratoryconfirmed positive test result, an FFD
policy violation may not be
administered. This laboratory
information must be treated as
potentially disqualifying FFD
information, unless the individual
subverts the screening process, in which
case a permanent denial of
authorization must be issued under
proposed § 26.610. To provide
assurance of testing effectiveness and
protections afforded to individuals
subject to the FFD program, proposed
§ 26.607(i) would require that an HHScertified laboratory must be used to
analyze the hair specimen, a forensic
toxicologist must review the licensee’s
or other entity’s hair screening process,
the test kit must be cleared by the FDA,
and hair screening must be conducted
in accordance with the HHS Guidelines.
The forensic toxicologist review would
be necessary if the panel of drug or drug
metabolites to be tested and their cutoffs
are not established by HHS or the NRC
for hair.
Proposed § 26.607(j) would allow the
use of portal area screening for drugs,
alcohol, or both. This provision would
result in a substantial contribution to a
licensee or other entity satisfying the
§ 26.23 performance objectives by
helping ensure that 100 percent of all
individuals who arrive at the NRClicensed facility to perform or direct
those duties and responsibilities or
maintain those types of access making
them subject to the FFD program are fit
for duty and deterred from arriving
onsite in a physiological condition that
may be adverse to safety and security.

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Additionally, screening could be
conducted when an individual exits the
NRC-licensed facility to provide
assurance that substance abuse had not
occurred on the site (see § 26.23(d)). The
screening device could be electronically
linked to temporarily prevent ingress or
egress and could automatically inform
licensee- or other entity-designated
officials of the portal area alarm. The
proposed requirement would enable the
licensee or other entity to use
innovative technologies to maintain
FFD program effectiveness when their
PMRP compels the licensee or other
entity to implement mitigative strategies
to maintain program effectiveness. The
use of portal screening technologies may
also represent cost savings because, for
NRC-licensed facilities that have small
staff sizes or are geographically remote,
passive drug and alcohol screening
technologies could be an innovative
alternative to a random testing program,
although the license or other entity
would need to request and receive an
exemption.
Proposed § 26.607(j) would also
provide that if the portal area screening
instrument detects a substance that
exceeds the instrument’s established
setpoint, the individual would be tested
with either a collection kit that must be
analyzed by an HHS-certified laboratory
or a POCTA. This situational screening
would be equivalent to a for-cause test.
The requirements would not allow an
individual to be rescreened by the portal
area screening instrument following an
initial screening detection that exceeded
an established setpoint in order to
prevent a subversion attempt. Similar to
other drug and alcohol testing
technologies enabled for use by subpart
M of part 26, a forensic toxicology
review would be required before using
passive screening technology to help
ensure the effectiveness of the
instrument by protecting against false
positive or negative screening results,
which would place an unwarranted
burden on the individual, licensee, or
other entity. These instruments and
alcohol screening devices, already in the
marketplace, may also be used to
determine true identity to facilitate
implementation of the FFD BOP, which
may be very practicable at facilities that
operate with small staff sizes.
Proposed § 26.607(k) would enable
the use of a blood specimen for drug,
alcohol, or other testing for certain
medical conditions as determined by
the licensee- or other entity-designated
MRO. This requirement would be
equivalent to current § 26.31(d)(5). The
use of a licensee- or other entitydesignated MRO and not one designated
by a third party, such as an MRO

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employed by an offsite specimen
collection facility, is important because
the MRO must be familiar with the
subpart M of part 26 requirements. To
help ensure testing effectiveness and
protect the worker, the blood test would
need to be conducted by a laboratory
that demonstrates compliance with
quality control requirements that are
comparable to those required for
certification by the HHS, such as a
hospital or clinic certified by the State,
Commonwealth, or territory.
Proposed § 26.607(l) would require
licensee and other entities to use a
Federal custody-and-control form (CCF)
approved by the OMB for the collection
and packaging of a hair, oral fluid, or
urine specimen. This proposed
requirement is based on the CCF
documentation requirements in current
subpart E of part 26 because subpart K
of part 26 does not require the use of a
CCF under § 26.117(e). Additionally,
when using a POCTA device, the
licensee or other entity would be
required to implement a licensee- or
other entity-approved and -maintained
procedure that ensures the reliability of
the tracking, handling, and storage of a
specimen from the point of specimen
collection to final disposition of the
specimen and the reliability of an
identification system to uniquely assign
the specimen to the donor. Both
requirements would help protect the
worker by helping ensure chain of
custody and by contributing to program
effectiveness.
Proposed § 26.607(m) would establish
requirements for the licensee- or other
entity-designated MRO. Section
26.607(m)(1) would be equivalent to
§ 26.405(g), however, the word
‘‘designated’’ would be added to the
first sentence to clarify that the MRO
would be designated by the licensee or
other entity, and not by a third party. As
stated with regard to proposed
§ 26.607(k), this change would clarify
that it is the licensee’s or other entity’s
responsibility, through their designated
MRO, to determine whether an
individual is fit for duty and
trustworthy and reliable. This would be
consistent with the description of FFD
program personnel in current § 26.31(b)
and help provide FFD program
effectiveness and protections to
individuals subject to the FFD program.
The paragraph was also modified from
§ 26.405(g) to address the
determinations of FFD policy violations
and fitness required by subpart H for a
part 53 licensee or other entity that
implements the FFD program described
in § 26.605(b).
Proposed § 26.607(m)(2) would help
ensure that MRO reviews are consistent

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with those MRO reviews conducted at
other NRC-licensed facilities subject to
part 26 and that the MRO maintains
knowledge of drug collection, testing
processes and procedures, and
evaluation of testing results.
The NRC also proposes that if an
MRO performed the duties and
responsibilities in §§ 26.185 and 26.187
for at least three continuous years in the
last 10 years prior to being hired or
contracted by the licensee or other
entity, then the MRO would not need to
repeat the initial training and
examination requirements. The basis for
3 years is that the MRO would have
experienced three annual cycles of
evaluating drug and alcohol test results,
contributed to the FFD annual report to
the NRC, experienced a refueling or
maintenance outage, understood the
duties and responsibilities of
individuals subject to the FFD program
to make informed determinations of
fitness, demonstrated a safety culture
that helps ensure FFD program
effectiveness, and been subject to NRC
inspection. The basis for 10 years is the
relatively long periods between
significant changes to part 26 and the
HHS Guidelines.
Proposed § 26.607(m)(3) would
require that the MRO attend a medicalor clinical-based training session on a
triennial basis. This proposal was
developed from Section 13.1 of the HHS
Guidelines for urine and oral fluid with
two substantial differences: the HHS
Guidelines state that ‘‘requalification
training,’’ including an exam, must be
conducted ‘‘at least every 5 years from
initial certification,’’ whereas the
proposed § 26.607(m)(3) would require a
training session every three years. The
proposed requirements are justified
because changes in societal drug use or
forensic toxicology could occur more
frequently than every 5 years, which
could compel MROs to attend training
in areas of forensic toxicology,
determinations of fitness, or other part
26 technical areas on a more frequent
periodicity than every 5 years to
improve their knowledge and expertise.
Proposed § 26.607(m)(4) would
require the MRO to evaluate drug testing
results by implementing the
requirements in § 26.185 or the HHS
Guidelines through the licensee’s or
other entity’s procedures. This
requirement would help ensure FFD
program effectiveness and enhance
consistency across the commercial
nuclear industry for the evaluation of
drug testing results. This also would
help protect individuals because they
would be subject to the same evaluation
criteria. If § 26.185 provides insufficient
information for an MRO to make a

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determination on a drug testing result
(including adulterant and discrepant
biological markers), the guidance issued
by a State agency in the state in which
the NRC-licensed facility is located,
Federal agency, or nationally recognized
MRO training and certification
organization may be used to inform an
MRO determination. This provision
would ensure that the MRO has the
flexibility to inform their evaluation of
the drug testing results and fitness
determination, if necessary, considering
the drug- and alcohol-related
flexibilities afforded in subpart M of
part 26.
The proposed requirement would also
state that an MRO need not review a
confirmed alcohol positive test result
determined by an EBT device under
§ 26.607(c)(3)(vi) and (vii), which are
equivalent to the current requirements
in §§ 26.101 and 26.103, respectively.
The results of an EBT device are precise
and accurate enough to support the
issuance of an FFD policy violation
without an MRO review of an EBT test
result if the instrument demonstrates
compliance with the requirements in
§ 26.91. The NRC acknowledges that
there are physiological conditions that
may cause an abnormally high blood
alcohol concentration, such as diabetes,
acid reflux, gastroesophageal reflux
disease, and perhaps certain diets (high
protein and low carbohydrates).
However, operating experience has not
demonstrated a compelling need to
require an MRO review of all EBT test
results. For consistency, a licensee or
other entity may elect to require its
MRO to review all EBT test results when
a donor communicates a testing concern
or physiological condition. If the donor
has a testing concern, the occurrence
could be appealed under the proposed
§ 26.613. If the donor presents a
physical condition to the MRO that may
have caused an elevated EBT test result,
the MRO may direct an alternative
testing process (see § 26.607(m)(5))
should it be medically necessary.
Proposed § 26.607(m)(5) would
require the licensee- or other entitydesignated MRO to determine and
approve the use of oral fluid or urine as
an alternative biological specimen when
the donor cannot provide a requested
specimen for testing. This proposed
requirement is equivalent to
§ 26.31(d)(5), which enables the use of
an alternative specimen collection if a
medical condition makes the collection
of the biological specimen difficult. This
determination and the retest must be
completed as soon as reasonably
practicable and documented to support
recordkeeping, auditing, and NRC
inspection.

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Proposed § 26.607(m)(6) would
require that the MRO review all
specimens screened or tested associated
with a drug-related FFD policy
violation. This includes POCTA, split
specimens, and all specimens taken to
resolve a discrepant condition, such as
a possible subversion attempt,
impairment without a known cause, or
a donor-requested or MRO-directed
retest. To resolve a discrepant
condition, the MRO is authorized to test
a specimen for a biological marker,
adulterants, or additional drugs. The
broad scope of this MRO evaluation
would be necessary because of the
variety of different screening and testing
methods that may have been associated
with the FFD policy violation. All
information learned from the conduct of
part 26 drug and alcohol screening and
testing should be used in the evaluation
of an individual’s trustworthiness and
reliability, issuance of a sanction, and
development of a follow-up treatment
and testing plan, if administered.
Proposed § 26.607(n) is equivalent to
current § 26.31(d)(6) and would
establish limits on the screening and
testing of biological specimens. This is
a protection consideration afforded to
individuals subject to the FFD program
and was not provided in subpart K of
part 26. This requirement states that
specimens collected under NRC
regulations may only be designated or
approved for screening and testing as
described in this part and may not be
used to conduct any other analysis or
test without the written permission of
the donor. Analyses and tests that may
not be conducted include, but are not
limited to, deoxyribonucleic acid (i.e.,
DNA) testing, serological typing, or any
other medical or genetic test used for
diagnostic or specimen identification
purposes.
The NRC proposes to require that no
biological specimens may be passively
sampled and analyzed in a manner
different than described in subpart M of
part 26 to ensure workers are protected
from non-consensual passive screening.
The subpart M framework enables
passive detection of drugs and alcohol,
whereas passive detection is not
afforded in subparts A through I, N, and
O of part 26.
Proposed § 26.607(o) is equivalent to
current §§ 26.31(b)(1)(iii)(A) and 26.89
and would require that all specimen
collections be conducted by a licenseeor other entity-designated and -trained
individual. For subpart M of part 26,
this would include onsite specimen
collections, except a collection by a
portal area screening instrument in
§ 26.607(j).

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Proposed § 26.608 would require
licensees and other entities to provide
FFD program training to individuals
subject to the FFD program. The
proposed performance-based § 26.608
requirement was developed from the
prescriptive training requirements in
current § 26.29 and modeled on current
§ 50.120 and the proposed requirements
in §§ 53.725 and 53.830 because there is
no training requirement in subpart K of
part 26.
Proposed § 26.608(a)(1) would require
an FFD training program that includes
the licensee’s or other entity’s FFD
policies and procedures, including
fatigue management, and the
individuals’ FFD program
responsibilities. Individuals who collect
specimens for testing or screening must
also be trained in specimen collector
duties and responsibilities, including, at
a minimum, specimen collection,
custody and control, identification and
response to subversion attempts, and
privacy. The fatigue management
training must include the knowledge
and abilities described in § 26.202(c).
For individuals specified in § 26.4, a
licensee or other entity of a commercial
nuclear plant would be required to use
a SAT as defined in proposed in
§ 53.725. These requirements are based
on requirements in § 26.29(a)(2), (3), (9),
and (10).
Proposed § 26.608(a)(2) would require
training on the BOP. This requirement
would be based on §§ 26.29(a)(8), (9),
and (10) and 26.33. The proposal would
require individuals to be trained in the
detection of behaviors or conditions
related to not only illegal drugs, as in
the current § 26.33 BOP requirements,
but also illicit drugs and substance
abuse onsite and offsite. Also, in
reference to impairment from fatigue or
any cause if left unattended, the phrase
in § 26.33, ‘‘may constitute a risk to
public health and safety or the common
defense and security,’’ would be
replaced in § 26.608(a)(2)(iii) with
‘‘could result in inattentiveness or
human errors,’’ because subpart M of
part 26 is focused, in part, on ensuring
individuals are fit for duty to safely and
competently perform or direct the
performance of assigned duties and
responsibilities.
Proposed § 26.608(a)(2)(iv) focuses on
training to inform individuals that they
are responsible for their own conduct,
as well as observing others. Specifically,
individuals would be trained to
recognize when they feel unable to
safely and competently perform
assigned duties and responsibilities or
act in a trustworthy and reliable
manner. The proposed training
requirement and the proposed reporting

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requirement in § 26.606(a)(5) are in the
interest of safety and security because
the individual is proactively
announcing that assistance may be
necessary. This would be consistent
with the performance objectives in
§ 26.23(b) and (c) where certain
behavior or stress conditions may be
indicative of an individual not being fit
for duty, trustworthy, and reliable.
Proposed § 26.608(a)(3) would help
ensure that individuals subject to the
FFD program understand that FFD
policy violations would result in an FFD
program sanction and that program
information learned or generated by
FFD program implementation would be
used to aide licensee or other entity
authorization determinations and be
shared, as requested, with other
licensees or other entities subject to
parts 26, 53, and 73. This proposed
requirement is equivalent to
§ 26.29(a)(1). Proposed § 26.608(a)(3)
would be a protection measure afforded
to individuals subject to the FFD
program because they would
understand that licensees and other
entities subject to parts 26, 53, and 73
would be informed of, in part, an
individual’s character, reputation, and
ability to follow policies, procedures,
and instructions to safely and
competently perform assigned duties
and responsibilities in a trustworthy
and reliable manner. Fitness-for-dutyrelated information would include drug
and alcohol testing results (not
quantitative testing values), issuance of
any sanctions, FFD-determinations
regarding trustworthiness and
reliability, testing programs, treatment,
and other remedial or corrective action.
Proposed § 26.608(b) would require
individuals be trained and receive a
trainee assessment before pre-access
testing and that refresher training and
trainee assessments be conducted
periodically thereafter. These
requirements would be equivalent to
§ 26.29(c)(1). However, § 26.608(b) was
developed from the SAT-based training
requirements in § 50.120 and training
elements from the annual training
requirements in § 26.29(c)(2). The term
‘‘systems approach to training’’ would
have the meaning in proposed
§ 53.725(c). A trainee assessment would
be the same as in currently required
SAT-based training programs.
Proposed § 26.608(c) would require
licensees and other entities to
periodically evaluate their FFD training
programs and revise them as
appropriate. This training focus is not
required by subpart K of part 26 or
§ 26.29 but is proposed to address the
flexibilities afforded in subpart M of

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part 26. This section would be
equivalent to § 50.120(b)(3).
Proposed § 26.609 would require the
implementation of a BOP. The proposed
requirement would be equivalent to that
in §§ 26.33 and 26.407, ‘‘Behavioral
observation,’’ and would apply during
construction, operation, and
decommissioning, if applicable. Because
subpart M of part 26 would apply
during decommissioning through a
licensee’s IMP, proposed § 26.609(a) and
(b) were developed, in part, from
proposed § 73.100(b)(9) and current
§§ 73.55(b)(9) and 73.56(f) to help
ensure consistency in the conduct of
behavioral observation whether
conducted for FFD or security purposes.
Under the FFD program, the purpose
of the BOP would be to help ensure that
individuals subject to the FFD program
are fit for duty and trustworthy and
reliable to perform or direct those duties
and responsibilities and maintain those
types of access that make the individual
subject to the FFD program. This
assurance is accomplished by requiring
each individual subject to subpart M of
part 26 to be subject to behavioral
observation, and by requiring all
individuals to perform behavioral
observation of others and report FFD
concerns to the licensee- or other entitydesignated representative(s). The intent
of the BOP requirement is not to require
that all individuals be observed at all
times by others; NRC-licensed operators,
maintenance professionals, security
officers, and others routinely perform
solo operations periodically throughout
the day. However, individuals must be
subject to observation while they are
performing or directing the performance
of duties and responsibilities or
maintaining the types of access making
them subject to the FFD program.
Observing behavior only at the
beginning of a work shift is not
sufficient to ascertain whether an
individual is fit for duty, trustworthy,
and reliable. Controlled substances may
have a delayed effect between use (e.g.,
ingestion) and the onset of physiological
or psychological effects, and fatigue
accumulates with time. Behavior must
be continually observed throughout the
work shift to detect any changes from
baseline human performance
characteristics, including mental or
physical health and mannerisms, or any
activities that may indicate that the
individual is not trustworthy and
reliable.
Proposed § 26.609(a) would differ
from §§ 26.33 and 26.407 in that it
would place the responsibility for
performing behavioral observation on
‘‘all individuals subject to this subpart,’’
rather than only those ‘‘individuals

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specified in § 26.4(f) [who] are
constructing or directing the
construction of safety- or securityrelated SSCs’’ in § 26.407 or
‘‘individuals who are trained under
§ 26.29 to detect behaviors’’ in § 26.33 to
improve clarity.
Proposed § 26.609(b) would require
all individuals subject to the FFD
program to report to the licensee- or
other entity-designated representative
any onsite or offsite behaviors or
activities by individuals subject to this
part that may constitute an
unreasonable risk to the safety or
security of the NRC-licensed facility or
SNM or may cause harm to others. The
NRC proposes this description of
reportable conduct because an
individual’s activities (e.g., use of illegal
substances) and communications (e.g.,
hate speech or threats of violence)
offsite are a direct indication of the
individual’s fitness, trustworthiness,
and reliability and must be evaluated as
to whether authorization should be
granted or maintained. Proposed
§ 26.609(b) would include a description
of this conduct instead of the § 26.33
undefined phrase, ‘‘FFD concerns,’’ to
enhance the clarity of the requirement.
This proposed BOP reporting
requirement would include any
information relating to character or
reputation of the individual indicating
that the individual cannot be trusted or
relied upon to perform those duties and
responsibilities or maintain access to
NRC-licensed facilities, SNM, or
sensitive information. This would better
align with the proposed § 73.120 BOP
requirement, which states that each
person subject to behavioral observation
must communicate to the licensee or
applicant observed behaviors or
activities of individuals that may
constitute an unreasonable risk to the
health and safety of the public and
common defense and security. Proposed
§ 26.609(a) and (b) were written broadly
to include offsite conduct that the
reporting individual considers serious
enough to call into question the
character or reputation of the subject
individual.
Proposed § 26.609(c) would require
that licensees and other entities perform
behavioral observation visually, inperson, and, when necessary, remotely
by live video and audible streaming and
capture. This requirement was
developed from the security observation
requirements in § 73.55(e)(7)(i)(B) and
(C), (h)(2)(v), and (i)(2) and (i)(5)(ii).
Conducting an in-person observation of
another individual is the preferred
method to ascertain whether the
observed individual can safely and
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and responsibilities. When in-person
observations are not feasible (e.g.,
during solo operations), the proposed
requirement would enable the use of
video monitoring. This is addressed, for
example, in proposed § 26.609(d)
regarding NRC-licensed operator
manipulation of reactor controls.
Additionally, certain duties (such as
maintenance activities performed by a
single worker outside of a control room)
may not present an opportunity for
video monitoring; in these situations,
behavioral observation should be
conducted on a sampling basis (i.e., a
planned observation of the work
activity) as outlined in a licensee’s or
other entity’s FFD program.
In situations involving small staff
sizes, facilities sited in geographically
remote locations, or both, additional
observers would enhance the
effectiveness of a BOP. Technological
developments in automated safety and
security systems may enable licensees
or other entities to reduce staff sizes to
10 to 40 percent of the staff size of an
LWR facility licensed under part 50 or
52. Smaller staff sizes may translate into
more solo operations, less teamwork,
fewer peer checks, or infrequent
management oversight of field activities,
leading to fewer behavioral
observations. Therefore, a licensee or
other entity would have fewer
opportunities to observe whether
individuals are fit for duty. Enabling
video and audible streaming and
capture to enhance the BOP would be
consistent with the security-related
behavioral observation requirement in
proposed § 73.120(c)(2)(ii), which
would also enable video conferencing or
other acceptable electronic means
promoting face-to-face interaction for
those individuals working remotely.
Proposed § 26.609(d) would require
that licensees or other entities perform
behavioral observation of NRC-licensed
operators who manipulate the controls
of any commercial nuclear plant
licensed under part 53, remotely by live
video and audible streaming capture for
those part 53 facilities where individual
task loading does not allow for the
effective conduct of behavior
observation in addition to assigned
operational tasks. The purpose of this
paragraph would be similar to that of
proposed § 26.609(c), where the
possibility of in-person observation is
significantly diminished because of solo
operations or because the facility may
only require a minimum staff size
onsite.
Proposed § 26.610 would be
equivalent to § 26.409, ‘‘Sanctions,’’ and
would require the licensee or other
entity to establish sanctions for FFD

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policy violations that, at a minimum,
prohibit the individuals specified in
§ 26.4 from being assigned to perform or
direct those duties and responsibilities
or maintaining authorization making
them subject to subpart M of part 26. To
be consistent with § 26.75, ‘‘Sanctions,’’
the severity of the sanction as described
in § 26.610 would escalate with the
number of occurrences and severity of
the FFD policy violation. The sanction
would be long enough to help deter
future FFD policy violations and
facilitate counseling and treatment
before the licensee reinstates the
individual’s access to the facility. The
NRC proposes this requirement because
the 14-day denial described in § 26.75
may not allow sufficient time for
counseling and treatment based on the
particular FFD policy violation.
Equivalent to § 26.75(c), proposed
§ 26.610 would also require a minimum
5-year denial of access to the NRClicensed facility for certain violations of
the FFD policy within the protected area
of a commercial nuclear plant and by an
individual or individuals who are the
operators of the conveyance to transport
or use formula quantities of strategic
SNM. Equivalent to § 26.75(b), proposed
§ 26.610 would require a permanent
denial of authorization be issued for any
subversion attempt.
Proposed § 26.611 would protect
information collected from FFD program
implementation and would be
equivalent to current § 26.411,
‘‘Protection of information.’’ The
protected information would include,
but not be limited to, privacy and
medical information. Section 26.611
would not include the § 26.411
requirement that FFD programs must
maintain and use the personal
information with the highest regard for
individual privacy because such a
requirement would be unnecessary in
light of the proposed § 26.611(a)
requirement that licensees and other
entities must establish and maintain a
system of files and procedures to
prevent unauthorized disclosure.
Proposed § 26.611(b), although
equivalent to § 26.411(b), would require
licensees and other entities to have all
individuals sign a consent to be subject
to the FFD program before subjecting
the individual to the FFD program (e.g.,
before being subject to a pre-access test
in § 26.607(b)(1), unlike § 26.411(b)).
The purpose of this proposal would be
to enhance protections afforded to
individuals subject to the FFD program
and their knowledge of, in part, why
they are subject to drug and alcohol
testing, behavioral observation,
information collection, MRO reviews,
and other FFD program elements. Like

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the consent required by § 26.411(b), the
consent would authorize disclosure of
the collected information. Consent
would not be needed for disclosures to
the individuals and entities specified in
§ 26.37(b)(1) through (b)(6), (b)(8), and
persons deciding matters under review
in proposed § 26.613, ‘‘Appeals
process.’’
Proposed § 26.613 would be
equivalent to § 26.413, ‘‘Review
process.’’ The proposed title was
changed to an appeal process to clarify
that § 26.613 would be the process
implemented when an individual elects
to appeal a licensee or other entity
determination that the individual had
violated the FFD policy. The proposal
would also require that the process
include a schedule for the completion of
the review of the determination that the
individual had violated the FFD policy.
The NRC proposes this requirement
because operating experience
demonstrates that workers may not be
protected from a continuous review
process that does not result in an
outcome.
Proposed § 26.615 would require
licensees and other entities to perform
audits of the FFD program. The
proposed section would be equivalent to
§ 26.415, ‘‘Audits.’’ Under proposed
§ 26.615(a), audits would be performed
at a frequency that ensures the FFD
program’s continuing effectiveness. This
would be particularly important for FFD
program elements that are not part of
the FFD PMRP required by § 26.603(d).
Corrective actions would be taken as
soon as reasonably practicable to resolve
any problems identified and preclude
recurrence. Proposed § 26.615(b) would
require the subject matter, scope, and
frequency of audits be revised as
necessary to improve or maintain
program performance based on findings
resulting from licensee or other entity
implementation of its FFD PMRP. These
requirements were developed from
appendix B to part 50, ‘‘Quality
Assurance Criteria for Nuclear Power
Plants and Fuel Reprocessing Plants’’;
criterion X, ‘‘Inspection’’; and criterion
XVIII, ‘‘Audits.’’
Proposed § 26.615(c) would be
equivalent to § 26.415(b) and would
enable licensees and other entities to
conduct joint audits or accept audits of
C/Vs so long as the audit addresses the
relevant services of the C/Vs.
Proposed § 26.615(d) would be
equivalent to § 26.415(c) by establishing
requirements for the auditing of HHScertified laboratories. Unlike § 26.415(c),
the proposal would not contain a
reference to the Department of
Transportation drug and alcohol testing
requirements. This would broaden the

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regulatory flexibility afforded to a
licensee or other entity in that they may
use an offsite collection or testing
facility that does not meet the
Department of Transportation
requirements.
Proposed § 26.615(d) would state that
licensees and other entities need not
audit an HHS-certified laboratory if the
licensee’s or other entity’s panel of
drugs and drug metabolites to be tested
is equivalent to the panel by which the
laboratory is certified by HHS or is
subject to the standards and procedures
for drug testing and evaluation used by
the laboratory under the HHS
Guidelines. The NRC would afford this
flexibility because the NRC is aware that
HHS desires to streamline changes in its
guidelines to its panel of drugs and drug
metabolites to be tested. Therefore, if a
licensee or other entity elects to
implement the HHS Guidelines in its
procedures and maintains the minimum
panel of drugs and drug metabolites to
be tested as required by subpart M of
part 26, a licensee or other entity may
still use (and not audit) the HHScertified laboratory because the
§ 26.603(e) change control process
would maintain FFD program
effectiveness.
To help ensure FFD program
effectiveness, § 26.615(d) would also
require that collection facility
procedures are comparable to those
required in subpart E of part 26,
including a proposed requirement that
the offsite facility’s specimen collection
and testing procedures are audited on a
biennial basis, which is also a
protection consideration afforded to
individuals subject to the FFD program.
Conducting this audit on a biennial
basis would be equivalent to that
required in § 26.41(b) and would help
ensure that the specimen collection
process at the facility remains effective.
Proposed § 26.617 would establish
recordkeeping and reporting
requirements equivalent to those in
current § 26.417. However, § 26.617
would require retention of records
pertaining to administration of the FFD
program and FFD performance data
required by § 26.717 until license
termination, which is based on current
§ 26.711(a) because § 26.417 does not
provide for a retention period.
Proposed § 26.617(b)(1) would be
identical to the reporting requirements
in § 26.417(b)(1) regarding the licensee’s
or other entity’s FFD program.
Proposed § 26.617(b)(2) would require
the reporting of annual (i.e., January
through December) program
performance information to the NRC
before March 1 of the following year.
This reporting would be equivalent to

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the annual program performance
requirement in § 26.417(b)(1), and the
March 1 due date is based on the
reporting deadline in § 26.717(e).
Licensees and other entities would be
required to report FFD performance
information using new NRC Forms 893,
‘‘Single FFD Policy Violation Form,’’
and 894, ‘‘10 CFR part 26, subpart M,
Annual Reporting Form for FFD
Performance Information.’’
Proposed § 26.617(c) would require
that FFD-related information be shared
within the commercial nuclear industry
when requested to support
authorization determinations. This
requirement would help individuals
seeking employment by another NRClicensed facility subject to subpart C of
part 26, complete their NRC-required
sanctions and licensee-administered or
-directed drug and/or alcohol abuse
treatment plans before the restoration of
authorization by a licensee or other
entity. Information sharing may also
enhance FFD program effectiveness
because FFD-related lessons learned
from, for example, substance testing,
subversion attempts, and laboratory and
MRO performance must be shared when
requested.
Proposed § 26.619 would require
licensees or other entities to establish a
process to evaluate individuals when
their fitness or trustworthiness and
reliability are in question. Section
26.619 would be equivalent to § 26.419,
‘‘Suitability and fitness
determinations,’’ but, unlike § 26.419,
would apply during the construction
and operation phases. Also, proposed
§ 26.619 would require that a suitability
or fitness determination conducted for
cause be conducted face-to-face. This
proposed requirement is based on
current § 26.189(c); however, unlike
§ 26.189(c), proposed § 26.619 would
not prohibit augmenting determinations
via electronic means of communication.
Instead, § 26.619 would explicitly
permit determinations to be performed
via electronic means, so long as those
determinations are supported by an
appropriately trained individual who is
present in-person with the individual
being assessed.
In considering the current restriction
on the use of electronic means of
communication for determinations of
fitness conducted for cause, the NRC
finds that since publication of the 2008
part 26 final rule, there have been
developments in using electronic means
of communication (i.e.,
‘‘videoconferencing’’) as an alternative
to conducting face-to-face interactions.
To address these considerations, the
NRC contracted the Pacific Northwest
National Laboratory (PNNL), DOE, to

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study whether a medical and mental
health assessment via electronic
communication could be an acceptable
alternative to an in-person, face-to-face
assessment.13 Based on this study, if
electronic means were to be used to
conduct a face-to-face assessment, an inperson element would still be integral to
the assessment process. However, under
certain circumstances, face-to-face
determinations and assessments
conducted as part of an FFD program for
an entity licensed under part 53 (i.e.,
those determinations and assessments
performed in accordance with § 26.619,
§ 26.207, or § 26.211) may be augmented
via electronic communications. Such
remotely conducted determinations and
assessments would be required to be
conducted with someone who is present
in-person with the individual being
assessed and who is trained in
accordance with the requirements of
either § 26.29 and § 26.203(c) or § 26.608
and § 26.202(c). Permitting the use of
electronic communications would help
ensure FFD program effectiveness,
especially in instances where the part
53 commercial nuclear plant is sited in
a geographically remote location or
when the facility has a small staff size.
D. Proposed Changes to Part 26, Subpart
N
Proposed § 26.709 would make the
recordkeeping and reporting
requirements in subpart N of part 26
applicable to licensees and other
entities of facilities licensed under part
53 that elect not to implement the
requirements in subpart M of part 26 or
elect to implement the requirements in
§ 26.605(b).
Proposed § 26.711(c) and (d) would be
amended to make these requirements
applicable to licensees or other entities
described in § 26.3(f). Section 26.711(c)
provides protection to individuals
subject to part 26 by enabling an
individual’s right to review FFD-related
information and correct any inaccurate
or incomplete information. Section
26.711(d) requires, in part, that any
FFD-related information shared with
other licensees or other entities is
correct and complete.
E. Proposed Changes to Part 26, Subpart
O
The vast majority of the proposed
changes to part 26 would be new or
revised substantive provisions that
would establish a regulatory obligation
or prohibition or would be conforming
edits to reflect the addition of part 53.
13 PNNL, Technical Letter Report, ‘‘The Use of
Electronic Communications to Perform
Determinations of Fitness,’’ dated August 2017.

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The only new provision that would not
be substantive, such that violation of it
would not result in a criminal penalty,
would be proposed § 26.601. Therefore,
the NRC proposes to add § 26.601 to the
list of regulations in § 26.825(b) to
which criminal sanctions do not apply.
10 CFR Part 50
A. Section 50.160: Emergency
Preparedness for Small Modular
Reactors, Non-Light-Water Reactors,
and Non-Power Production or
Utilization Facilities
This proposed rule would revise
§ 50.160(b)(3) and (c)(2) to make that
section applicable to applicants and
licensees under part 53. Section 50.160
provides an alternative to other part 50
emergency preparedness requirements
focused on large light-water reactors to
provide an optional emergency
preparedness framework specifically for
small modular reactors (SMRs) and
other new technologies. These
alternative emergency preparedness
requirements adopt a performancebased, technology-inclusive, riskinformed, and consequence-oriented
approach. Commercial nuclear reactor
applicants complying with § 50.160
would be required to submit as part of
the application the analysis used to
determine whether the criteria in
§ 53.1109(g)(2)(i)(A) and (B) are met
and, if they are met, the size of the
plume exposure pathway emergency
planning zone (EPZ). An EPZ bounds
the area surrounding a facility within
which detailed planning is needed to
implement predetermined, prompt
protective actions. The criterion in
proposed § 53.1109(g)(2)(i)(A) is that
public dose, as defined in § 20.1003, is
projected to exceed 10 mSv (1 rem)
TEDE over 96 hours from the release of
radioactive materials from the facility
considering accident likelihood and
source term, timing of the accident
sequence, and meteorology. The
criterion in proposed
§ 53.1109(g)(2)(i)(B) is that predetermined, prompt protective measures
are necessary. These are the same
criteria that are in § 50.33(g)(2)(i)(A) and
(B) and are used to assess the need for
and size of an EPZ in applications under
parts 50 and 52.
Applicants choosing to comply with
§ 50.160 must determine the
radiological releases from the facility
that are evaluated in the determination
of the plume exposure pathway EPZ.
Consistent with other Federal guidelines
such as the Federal Emergency
Management Agency ‘‘Radiological
Emergency Preparedness Program
Manual,’’ issued in 2023, and the

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Environmental Protection Agency ‘‘PAG
Manual: Protective Action Guides and
Planning Guidance for Radiological
Incidents,’’ issued in 2017, applicants
should consider quantitative and
qualitative information on the potential
radiological releases that make up the
spectrum of accidents used to develop
the basis for the applicant’s site-specific
EPZ. This information is derived from
the licensing basis. The NRC plans to
update the risk-informed approach in
RG 1.242 for part 53 while maintaining
its flexibility for using information
already developed and available in
licensing basis documents, including
PRA results, deterministic dose
quantities, accident timing, target set
analyses, mitigation capabilities, and
site-specific factors such as
meteorology.
In its safety analysis report, the
applicant would describe the LBEs
relevant to the facility and would
consider these LBEs as candidates for
the spectrum of accidents used to
develop the site-specific EPZ. The LBEs
assessed include a wide range of events
that are appropriate for considering in
the facility’s emergency preparedness
and response planning. In addition,
§ 50.160(b)(1)(iv)(A)(2) requires
licensees to be capable of implementing
their approved emergency response plan
in conjunction with their safeguards
contingency plan. Radiological sabotage
events are typically factored into EPZ
determinations by considering
consequences to be bounded by LBEs
and by crediting protection against the
DBT in reducing the likelihood of a
release.
The provisions in proposed
§ 53.860(a) provide an alternative to
applicants and licensees by not
requiring them to protect against the
DBT of radiological sabotage in
accordance with §§ 73.55 and 73.100 if
they can demonstrate that the
consequences from unmitigated
radiological sabotage events are below
the safety criteria in proposed § 53.210.
The deployment of some commercial
nuclear plants under part 53 may
involve new scenarios where the source
terms and consequences of sabotagerelated events are not bounded by the
consequences of the unlikely and very
unlikely event sequences analyzed
under subpart C. Accordingly, the NRC
plans to develop guidance for part 53
applicants and licensees choosing to
comply with the alternative emergency
preparedness requirements in § 50.160
to address this new class of reactors. In
Section VI of this document, the NRC is
asking for stakeholder feedback on the
clarity of the regulations and guidance
for various scenarios that might arise in

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implementing graded approaches for
security and emergency planning for
some commercial nuclear plant designs.
B. Appendix B to Part 50: Quality
Assurance Criteria for Nuclear Power
Plants and Fuel Reprocessing Plants
Appendix B to part 50 would be
amended to make it applicable to
applicants and licensees under part 53.
This results in the need for some
revisions to recognize differences in
terminology between parts 50 and 53.
Namely, the term ‘‘design bases,’’ which
is defined in § 50.2, is not used in part
53. For this reason, text is added in both
Section III, ‘‘Design Control,’’ and
Section IV, ‘‘Procurement Document
Control,’’ to refer to ‘‘functional design
criteria, as defined in § 53.020,’’ as the
part 53 equivalent of the term ‘‘design
bases.’’
10 CFR Part 73
A. Section 73.100: Technology-Inclusive
Requirements for Physical Protection of
Licensed Activities at Commercial
Nuclear Plants Against Radiological
Sabotage
Proposed § 73.100 would provide a
performance-based regulatory
framework for the design,
implementation, and maintenance of a
physical protection program and
security organization for certain
commercial nuclear plants licensed
under part 53. The current § 73.55
physical security requirements for
nuclear power reactors licensed under
part 50 and part 52 use a combination
of performance criteria (e.g.,
§ 73.55(b)(1) through (3)) and numerous
prescriptive requirements developed to
achieve performance objectives (e.g.,
§ 73.55(k)(5)(ii)). By contrast, in the
proposed performance-based approach
to physical security for part 53,
performance objectives and
requirements would be the primary
bases for regulatory decision-making,
giving the licensee the flexibility to
determine how to demonstrate
compliance with the established
performance criteria for an effective
physical protection program. This
proposed physical protection program
would provide an optional pathway for
licensees that elect not to demonstrate
compliance with the provisions in
§ 73.55 and do not satisfy the criterion
as described in proposed § 53.860(a)(2).
This proposed physical protection
program would provide that activities
involving SNM are not inimical to the
common defense and security and do
not constitute an unreasonable risk to
the public health and safety.

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Section 73.100(a) would require each
part 53 licensee that elects to
demonstrate compliance with this
section rather than § 73.55 to implement
the requirements therein through its
physical security plan, training and
qualification plan, safeguards
contingency plan, and cybersecurity
plan (referred to collectively hereafter as
‘‘security plans’’) prior to initial fuel
load into the reactor (or, for a fueled
manufactured reactor, before initiating
the physical removal of any one of the
independent physical mechanisms to
prevent criticality required under
§ 53.620(d)(1)). The security plans
would need to identify, describe, and
account for site-specific conditions that
affect the licensee’s capability to satisfy
the requirements of § 73.100. Based on
experience from recent new reactor
licensing reviews, the NRC recognizes
that licensees may seek to receive
unirradiated fuel onsite before carrying
out the security requirements in
§ 73.100. However, these security
requirements would have to be
implemented at some point before
reactor operation to address the
increased risk arising from irradiated
fuel onsite. This proposed rule would
make clear that part 53 applicants and
licensees using § 73.100 may bring
unirradiated nuclear fuel onsite and
protect it in accordance with the NRC’s
requirements for physical protection of
SNM of moderate and low strategic
significance under § 73.67 until initial
fuel load into the reactor (or, for a fueled
manufactured reactor, until initiating
the physical removal of any one of the
independent physical mechanisms to
prevent criticality required under
§ 53.620(d)(1)).
Section 73.100(b) would outline the
general performance objective and
design requirements of the licensee
physical protection program. A
licensee’s program would be required to
provide protection against any
deliberate act within the DBT of
radiological sabotage, including spent
fuel sabotage, which could directly or
indirectly endanger the public health
and safety by exposure to radiation. The
physical protection program is
supported by the AA program,
cybersecurity program, and IMP to
demonstrate compliance with the
general performance objective of
§ 73.100(b).
Section 73.100(b)(2) was developed,
in part, from § 73.55(b)(3). To satisfy the
general performance objective of
§ 73.100(b)(1), the physical protection
program would need to protect against
the DBT of radiological sabotage. The
existing fleet of LWR satisfies this
objective by preventing significant core

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damage and spent fuel sabotage. Some
non-LWR reactor licensees’ physical
protection programs may be designed to
prevent a significant release of
radionuclides from any source.
Therefore, the proposed performance
objective would focus on radiological
sabotage in general, rather than a
specific focus on core damage or spent
fuel sabotage, to be technology inclusive
and allow for flexibility for different
reactor technologies.
Under the proposed § 73.100(b)(2)(ii),
licensees must provide defense in depth
in achieving performance requirements
through the integration of engineered
systems, administrative controls, and
management measures. This
requirement would apply defense-indepth concepts as part of the physical
protection program to ensure the
capability to demonstrate compliance
with the performance objective of the
proposed § 73.100(b)(1) is maintained in
the changing threat environment. The
defense-in-depth philosophy applies to
measures against intentional acts as
required by § 73.100(b), and the designs
of physical security systems should
employ defense in depth through
systems diversity, independence, and
separation under § 73.100(b)(2). The
most common defense-in-depth
measures apply concepts of
redundancy, diversity, independence,
and safety margin to ensure systems
reliability and availability. The defensein-depth philosophy applies to the
design of a physical protection program,
which integrates engineered controls
and administrative controls, to provide
protection against the DBT for
radiological sabotage.
Section 73.100(b)(3) would require
the physical protection program to be
designed and implemented to achieve
and maintain the reliability and
availability of SSCs required for
demonstrating compliance with
specified performance requirements.
These physical protection performance
requirements were informed by
§ 73.55(b) and the Commission’s
Advanced Reactor Policy Statement.
The performance objective of
protecting against the DBT of
radiological sabotage is achieved by the
design and implementation of the
physical protection program,
maintained at all times, with the
following required performance
capabilities proposed in the provisions
in § 73.100(b)(3): intrusion detection,
intrusion assessment, security
communication, security response,
protecting against land and waterborne
vehicle bomb assaults, and access
control portals. The physical protection
program must maintain the reliability

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and availability of SSCs relied upon for
demonstrating compliance with the
performance requirements. The terms
‘‘reliability and availability’’ are
intended to describe defense in depth in
a performance-based manner and would
be critical elements for demonstrating
compliance with the proposed
requirement for protection against the
DBT of radiological sabotage as
described in the proposed
§ 73.100(b)(2).
The first element, ‘‘intrusion
detection,’’ would be provided through
the use of detection equipment, patrols,
access controls, and other program
elements and would provide
notification to the licensee that a
potential threat is present and where the
threat is located.
The second element, ‘‘intrusion
assessment,’’ would provide a
mechanism through which the licensee
would identify the nature of the threat
detected. This would be accomplished
through the use of video equipment,
patrols, and other program elements that
would provide the licensee with timely
information about the threat for use in
determining how to respond.
The third element, ‘‘security
communication,’’ would provide a
mechanism through which the licensee
would communicate the necessary
information to the response force to
ensure effectiveness of the physical
protection program. This would be
accomplished through the redundant,
independent, and diverse design of
physical security and/or plant SSCs
relied on for onsite and offsite security
communications. The continuity and
integrity of communications should
account for the DBT’s ability to affect
the reliability and availability of
communications.
The fourth element, ‘‘security
response,’’ would provide a mechanism
through which the licensee would be
capable of timely security response to
interdict and neutralize threats up to
and including the DBT of radiological
sabotage. The security response may
include the use of onsite armed
responders, law enforcement responders
(local, State, or Federal), or other offsite
armed responders (e.g., licensee
proprietary or contract security
personnel who are positioned offsite), or
a combination thereof, as appropriate.14
14 The NRC’s security regulations for commercial
nuclear power reactors have historically considered
onsite armed responders to be the only acceptable
method for interdicting and neutralizing threats up
to and including the DBT of radiological sabotage.
The proposed rule would permit advanced power
reactor licensees to use any interdiction and
neutralization method, which would be an
extension of the Commission’s position in SRM–

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The licensee must provide protection
against any element of the DBT, to
include those that do not rise to the full
capability of the DBT. Structures,
systems, and components relied on to
provide delay functions must be
designed to provide for timely response
to adversary attacks with adequate
defense in depth. Delay would allow the
licensee to take necessary actions to
counter any attempt by the threat to
advance towards the protected target or
target set element. The overall response
objective would be to place the threat in
a condition from which the threat no
longer has the potential for, or
capability of, doing harm to the
protected target.
The fifth element, ‘‘protecting against
land and waterborne vehicle bomb
assaults,’’ would provide a mechanism
through which the licensee would be
capable of protecting the plant against
the DBT vehicle bomb assault. The
methods that are relied on to protect
against a DBT land vehicle and
waterborne vehicle bomb assault must
be designed to protect the reactor
building, structures containing safety or
security related systems, and
components from explosive effects.
The sixth element, ‘‘access control
portals,’’ would provide a mechanism
through which the licensee would be
capable of detecting and denying
unauthorized access to persons and
pass-through of contraband materials
(e.g., weapons, incendiaries, explosives)
to protected areas. Integrity of the access
control system is maintained through
licensee oversight and ensures that
attempts to circumvent or bypass the
established process will be detected and
access denied.
The proposed performance
requirements would permit the
applicant or licensee to determine how
to design the physical protection
program to protect the plant against the
DBT of radiological sabotage without
SECY–17–0100, ‘‘Security Baseline Inspection
Program Assessment Results and Recommendations
for Program Efficiencies,’’ dated October 8, 2018,
and SRM–SECY–20–0070, ‘‘Technical Evaluation of
the Security Bounding Time Concept for Operating
Nuclear Power Plants,’’ dated June 6, 2024. Under
the proposed rule, a licensee would retain the
responsibility to detect, assess, interdict, and
neutralize threats up to and including the DBT of
radiological sabotage, but would be able to rely on
law enforcement or other offsite armed responders
as a method for fulfilling the required interdiction
and neutralization capabilities. For licensees that
choose to rely on law enforcement to fulfill these
capabilities, the proposed rule would not create any
NRC regulatory jurisdiction over, or requirements
for, law enforcement. In SRM–SECY–23–0021,
‘‘Proposed Rule: Risk-Informed, TechnologyInclusive Regulatory Framework for Advanced
Reactors (RIN 3150–AK31),’’ dated March 4, 2024,
the Commission approved a similar approach to
defend against radiological sabotage.

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prescriptive requirements such as those
currently found in § 73.55. DG–5076,
‘‘Guidance for Technology Inclusive
Requirements for Physical Protection of
Licensed Activities at Commercial
Nuclear Plants,’’ has been developed by
the NRC to describe one acceptable
approach to demonstrate compliance
with requirements proposed in § 73.100.
Section 73.100(b)(4) would require
the licensee to identify target sets in
accordance with § 73.55(f). For nonLWR and SMRs, target sets would be
defined in DG–5071, ‘‘Target Set
Identification and Development for
Nuclear Power Plants,’’ as the minimum
combination of equipment, operator
actions, and/or structures that, if all are
prevented from performing their
intended safety function or prevented
from being accomplished, barring
extraordinary actions by plant
operations, would likely result in a
significant release of radionuclides from
any source (e.g., a release to the
environment exceeding that analyzed in
the DBA licensing basis).
Section 73.100(b)(5) would require
that each licensee perform a site-specific
analysis for the purpose of identifying
and analyzing site-specific conditions
that affect the design of the onsite
physical protection program.
Section 73.100(b)(6) would require
licensees to implement a performance
evaluation program, which would
ensure that a licensee will periodically
test and evaluate the effectiveness of the
physical protection program to protect
against the DBT. This program would
ensure that licensees are able to
demonstrate that the physical protection
program satisfies the response
requirements of § 73.100 and that the
site’s protective strategy effectively
protects against the DBT. Licensee
performance evaluations would include
methods to assess, test, and challenge
the integration of the physical
protection programs functions and
demonstrate the effectiveness of security
plans, licensee protective strategy, and
implementing procedures in accordance
with § 73.100(g).
Section 73.100(b)(7) would require
licensees to implement an AA program
in accordance with § 73.56. Section
73.100(b)(8) would require licensees to
establish, maintain, and implement
protection against a cyberattack based
on either the proposed cybersecurity
program described in § 73.110 or the
program described in existing § 73.54.
Section 73.100(b)(9) would require an
IMP that monitors the initial and
continuing trustworthiness and
reliability of individuals granted or
retaining unescorted access or
unescorted AA to a protected or vital

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area. The IMP must also implement
defense-in-depth methodologies to
minimize the potential for an insider
(active, passive, or both) to adversely
affect the licensee’s capability to protect
against radiological sabotage. Because
no one element of the AA program, FFD
program, cybersecurity program, or
physical protection program, would, by
itself, provide the level of protection
against the insider necessary to
demonstrate compliance with the
performance objective of the proposed
§ 73.100(b), the effective integration of
these programs is a necessary
requirement to achieve defense in depth
against the potential insider.
Section 73.100(b)(10) would require
that the licensee have the capability to
track, trend, correct, and prevent
recurrence of failures and deficiencies
in the implementation of the
requirements of this section. Section
73.100(b)(11) would require the
coordination of the security plans and
associated procedures with other onsite
plans to manage the safety and security
interface during normal or emergency
operations.
Section 73.100(c) was developed from
§ 73.55(c)(7), ‘‘Security implementing
procedures,’’ and § 73.55(d), ‘‘Security
organization,’’ and would outline the
requirements for the composition,
equipping, and training of the security
organization. The purpose of the
security organization is to effectively
implement the physical protection
program. Individuals assigned to
perform physical protection or
contingency response duties must be
trained, equipped, and qualified to
perform assigned duties and
responsibilities.
Section 73.100(d) would establish a
performance requirement for searches of
personnel, vehicles, and materials for
the protection against radiological
sabotage. The requirement describes
broad categories of material (explosives,
firearms, incendiary devices, etc.) to be
detected and prevented from entry into
the protected area; specific items that
will be prohibited would not be
prescribed in the regulation but will be
stated in the licensee security plans
with detailed descriptions being
identified in implementation
procedures.
Section 73.100(e) would require a
training and qualification program,
described in the training and
qualification plan, that ensures
personnel are able to effectively perform
their assigned security-related job
duties. This high-level requirement
would allow flexibility in how the
licensee chooses to train its security
personnel. One method for

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accomplishing this requirement would
be to provide a training and
qualification program that would be
equivalent to appendix B to part 73.
Section 73.100(f) would require
periodic security reviews of the physical
protection program to ensure effective
implementation of the program by
independent individuals. The
evaluation process would provide a
systematized approach for assessing the
physical protection program as a basis
for further development and
improvement. Program reviews should
be designed to ensure that the physical
protection program maintains
effectiveness and demonstrates
compliance with NRC requirements.
Section 73.100(f)(1) was developed from
§ 73.55(m) and would require review of
each element of the physical protection
program. Section 73.100(f)(2) would
require licensees to perform selfassessments of physical protection
program functions to ensure that the
capability to detect, assess, interdict,
and neutralize the DBT of radiological
sabotage is maintained. Section
73.100(f)(3) would require an audit of
the effectiveness of the physical
protection program; security plans;
implementing procedures; cybersecurity
programs; management of the safety/
security interface activities; the testing,
maintenance, and calibration program;
and response commitments by local,
State, and Federal law enforcement
authorities. Section 73.100(f)(4) would
require that results and
recommendations, management
findings, and any actions taken be
documented and maintained to be
available for inspection by the NRC.
These reviews are independent of the
ongoing performance evaluations
described in § 73.100(b)(6) and (g).
Section 73.100(g) would require that
licensee performance evaluations,
described in § 73.100(b)(6), include
methods appropriate and necessary to
assess, test, and challenge the
integration of the physical protection
program’s functions to protect against
the DBT. The performance evaluations
must also address the licensee’s
measures to protect against cyberattacks,
in accordance with the required
cybersecurity plan, and engineered
systems designed to protect against the
DBT standalone ground vehicle bomb
attack.
Section 73.100(h) would establish
performance requirements for
maintaining security SSCs relied on to
perform security functions to protect
against the DBT. It would require that
corrective actions and compensatory
measures be taken by a licensee in
response to a degradation of security

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equipment or failure of the equipment
to perform its intended functions. The
licensee would be required to maintain
the SSCs described in its design and
licensing basis to ensure that they are
reliable and available.
Section 73.100(i) would establish
requirements for the suspension of
security measures in response to
emergency and extraordinary
conditions. The requirements of this
paragraph, which were developed from
§ 73.55(p), would be intended to
provide flexibility to a licensee for
taking reasonable actions that depart
from a security plan in an emergency
when such actions are immediately
needed to protect the public health and
safety and no action consistent with
license conditions and TS that can
provide adequate or equivalent
protection is immediately apparent in
accordance with proposed § 53.740(h).
Section 73.100(j) would establish
requirements regarding the inspection,
retention and maintenance of records
required to be kept by the NRC
regulations, orders, or license
conditions. These proposed
requirements are developed from
§ 73.55(q).
B. Section 73.110: Technology-Inclusive
Requirements for Protection of Digital
Computer and Communication Systems
and Networks
Section 53.860 would require that a
licensee establish, implement, and
maintain a cybersecurity program in
accordance with § 73.54 or § 73.110.
Section 73.110 would establish
requirements for the development and
maintenance of a cybersecurity program
for commercial nuclear plants licensed
under part 53. This proposed section
would implement a graded approach to
determine the level of cybersecurity
protection required for digital
computers, communication systems,
and networks. The proposed new
section is informed by: (1) the operating
experience from power reactors and fuel
cycle facilities; and (2) the existing
§ 73.54 framework, which addresses
some of the basic issues for
cybersecurity regardless of the type of
reactor. Differences between the § 73.54
requirements and those proposed in
§ 73.110 are primarily based on the
implementation of a consequence-based
approach to cybersecurity that provides
flexibility to accommodate the wide
range of reactor technologies to be
assessed by the NRC. A graded approach
based on consequences is intended to
account for the differing risk levels
among reactor technologies.
Specifically, the proposed new section
would require licensees to demonstrate

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protection against cyberattacks in a
manner that is commensurate with the
potential consequences from those
attacks.
Under proposed § 73.110(a), licensees
would need to ensure that digital
computer and communications systems
are adequately protected against a
potential cyberattack that would result
in: (1) a scenario where the cyberattack
leads to offsite radiation doses that
would endanger public health and
safety (i.e., the resulting consequence
exceeds the reference dose values in
§ 53.210); or (2) a scenario where the
cyberattack adversely impacts the
physical security digital assets used by
the licensee to prevent unauthorized
removal of material or radiological
sabotage. Security digital assets would
include those used for nuclear MC&A.
The proposed § 73.110(b) would
require licensees to protect the
communication system and networks
associated with the functions described
in § 73.110(a)(1) and (a)(2) from
cyberattacks. To accomplish this, the
licensee would establish, implement,
and maintain a cybersecurity program
for protecting digital assets within the
scope of § 73.110 that would make use
of risk insights, including threat
information, and would consider the
resulting level of consequences of the
threats. If the outcome of the assessment
by the licensee under § 73.110(b)(1)
revealed that a potential cyberattack
would not compromise any digital
assets that support safety and security
functions, and thus would not result in
the consequences listed in § 73.110(a)
(e.g., would not exceed the reference
dose values), then only a narrow set of
the cybersecurity program requirements
in § 73.110(d) and (e) would apply. For
example, the licensee would only need
to develop a cybersecurity program that
implements the requirements dealing
with:
• Analyzing modifications of any
asset before implementation to see if
they demonstrate compliance with the
potential consequences in § 73.110(a);
• Ensuring employees and contractors
are aware of cybersecurity requirements
and have some level of cybersecurity
training;
• Evaluating and managing
cybersecurity risks to the plant;
• Reviewing the cybersecurity plan
for any required changes; and,
• Retaining records of the
cybersecurity plan along with any plan
changes.
Section 73.110(c) through (e) were
developed from § 73.54(a)(2), and (c)
through (h), respectively.
The proposed requirements would
address the need for the licensee to

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develop a cybersecurity program that
implements a defense-in-depth
protective strategy as required by
proposed section § 73.110(d)(2). A
defense-in-depth protective strategy for
cybersecurity is represented by
collections of complementary and
redundant security controls that
establish multiple layers of protection to
safeguard critical digital assets. Under a
defense-in-depth protective strategy, the
failure of a single protective strategy or
security control should not result in the
compromise of safety and security
functions.
C. Section 73.120: Access Authorization
Program for Commercial Nuclear Plants
Section 73.120 would address AA for
certain commercial nuclear plants
licensed under part 53. The proposed
language in § 73.120 would provide an
alternate approach to the existing
framework for AA under §§ 73.55,
73.56, and 73.57, commensurate with
risk and consequences to public health
and safety. It would be available to part
53 applicants and licensees who
demonstrate in an analysis that the
offsite consequences of a DBE satisfy the
criterion defined in § 53.860(a)(2)(i) (i.e.,
would not exceed the offsite dose values
in § 53.210(b)). The proposed
requirements in § 73.120 would be
similar to the existing AA program
elements for those NRC licensed
facilities issued additional security
measures (ASMs) orders and for
materials licensees under § 37.21.
Applicants not satisfying the criterion
would need to establish, implement,
and maintain a full AA program,
including an IMP, in accordance with
§ 73.56.
Proposed § 73.120(a) would be based
on an applicant satisfying the eligibility
criterion in § 53.860(a)(2)(i). Section
73.120(b) would identify the categories
of individuals who would be subject to
an AA program in accordance with this
section. The applicability statement in
§ 73.120(b)(1)(i) would encompass
individuals whom the licensee intends
to grant unescorted access to the
facilities’ most sensitive areas,
consistent with § 73.56(b)(1)(i) for
power reactors and the ASM orders and
license conditions issued to any NRC
licensed facility or material licensee.
Sections 73.120(b)(1)(ii) through (iv)
would be consistent with
§ 73.56(b)(1)(ii) through (iv),
respectively. The program would
include individuals who may be onsite
or offsite (e.g., remote operators or
information technology staff) and have
virtual access to important plant
operational and communication systems
based upon assigned duties and

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responsibilities. An individual who has
remote access to plant equipment and
communication systems may have
trusted privileges greater than the
personnel at the plant site. Section
73.120(b)(1)(iii) would state that offsite
law enforcement personnel on official
duty would not be subject to the
licensee AA program.
Section 73.120(c) would provide
general performance objectives and
requirements largely consistent with the
AA program requirements for nuclear
power reactors under § 73.56 and would
provide licensees and applicants the
flexibility in establishing their AA
program to demonstrate compliance
with various performance objectives.
Section 73.120(c)(1) would include
background investigation requirements
consistent with § 37.25, as well as ASMs
and license conditions that are applied
to non-power reactor licensees.
Background investigations include
important elements to establish the
trustworthiness and reliability of an
individual, such that they do not
constitute an unreasonable risk to
public health and safety or the common
defense and security. These include the
following: (1) personal history
disclosure, (2) verification of true
identity, (3) employment history
evaluation, (4) unemployment/military
service/education, (5) credit history
evaluation, (6) character and reputation
evaluation, and (7) Federal Bureau of
Investigation criminal history record
check.
Section § 73.120(c)(2) would establish
behavioral observation requirements,
which are an awareness initiative for
recognizing behaviors adverse to the
safe operation and security of the
facility through observing the behavior
of others in the workplace and reporting
aberrant behavior or changes in
behavior that might reflect negatively on
an individual’s trustworthiness or
reliability. Maintaining behavioral
observation would assist and/or
improve worker safety and reduce the
risk of an insider threat. This proposed
requirement in § 73.120(c)(2) would be
a scaled version of the full BOP required
under § 73.56(f).
Section § 73.120(c)(2) would provide
licensees greater flexibility to
implement behavioral observation
options for individuals granted
unescorted access to the commercial
nuclear plant’s protected area. Such
options on reporting questionable
behavior may include a program similar
to the Department of Homeland
Security’s program, ‘‘If you see
something, say something,’’ or to a
corporate behavioral awareness
program. Commensurate with the

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potential lower safety and security risks
of a commercial nuclear plant that
meets the criterion in § 53.860(a)(2)(i),
§ 73.120(c)(2) would not require the
establishment of a comprehensive
training program for behavioral
observation (i.e., initial and refresher
training including knowledge checks) as
required for power reactors under
§ 73.56 and part 26. Under
§ 73.120(c)(2)(ii), behavioral observation
would be able to be performed in-person
or remotely by video, and identified
behavior of concern would need to be
reported to plant supervision. The
remote access alternative to face-to-face
interactions provides substantial
flexibility for licensees and applicants.
Any video conferencing or other
acceptable electronic means promoting
face-to-face interaction for those
individuals working remotely would
demonstrate compliance with this
regulation.
Section 73.120(c)(3) captures and
maintains the self-reporting of legal
actions as an essential performance
element to enhance the licensee’s
behavioral observation initiative similar
to the current requirements under
§ 73.56(g), assuring that personnel who
are granted and who maintain
unescorted access are trustworthy and
reliable.
Section 73.120(c)(4) would provide a
scalable approach for granting and
maintaining unescorted access. One
component not included from § 73.56 is
the need for a psychological assessment
and reassessment under § 73.56(e) for
granting unescorted access and
§ 73.56(i)(v)(B) for individuals who
perform one or more of the job functions
described in § 73.120(b)(1)(ii) for
maintaining unescorted access.
Moreover, the requirement would
permit criminal history updates to be
completed within 10 years of the last
review, compared to the three- or fiveyear reinvestigation periodicity for
personnel at an operating commercial
nuclear plant. In addition, no credit
check re-evaluation would be required
for these individuals.
The continued need to maintain
unescorted access would be evaluated
on an annual basis by the reviewing
official. Guidance in DG–5074, ‘‘Access
Authorization Program for Commercial
Nuclear Plants,’’ would specify that this
evaluation should be based on a
compilation of personnel interactions as
described in the licensee’s or applicant’s
policy and procedures for behavioral
observation and the maintenance of an
approved AA list.
Section 73.120(c)(5) would require
licensees and applicants to determine
when a person no longer requires the

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need for unescorted access or no longer
satisfies the AA requirement found
within this section. Guidance in DG–
5074 would further explain that
licensees have the flexibility to
terminate unescorted access to specific
areas of the site if individuals lack the
continued need for that access to
perform their duties and
responsibilities.
Section 73.120(c)(6) would be
consistent with the purpose of § 37.23(e)
and would include the individual’s
right to correct and complete
information as required under
§ 37.23(g). The section would include a
requirement for designating a reviewing
official. The language would provide
clarity regarding the roles and
responsibility of a reviewing official,
who would be the only individual
authorized to make unescorted access
determinations.
Section 73.120(c)(7) would align with
the corresponding requirements under
§ 37.23(f), and § 73.120(c)(8) would
align with the corresponding
requirements under § 37.31. These
requirements would encompass the
roles and responsibilities for licensees,
applicants, and if applicable, the
contractor/vendors to establish,
implement, and maintain a system of
files and records to ensure personal
information is not disclosed to
unauthorized persons.
Section 73.120(c)(9) would align with
the requirements of § 37.33. Section
73.120(c)(10) would require licensees,
applicants, and contractors or vendors
to maintain the records that are required
by the regulations in this section and
retain them for a period of 3 years after
the record is superseded or no longer
needed. The record retention period of
three years would be consistent with
§ 37.23(h), contrasting with the five-year
retention period under § 73.56(o).
Records maintained in any database(s)
would need to be available for NRC
review, consistent with the
requirements found under
§ 73.56(o)(6)(ii).
VI. Specific Requests for Comments
The NRC is seeking advice and
recommendations from the public on
this proposed rule. We are particularly
interested in comments and supporting
rationale from the public on the
following:
Part 26—Fitness for Duty Program
1. The proposed rule under
§ 26.603(c) would enable a licensee or
other entity to implement an FFD
program under proposed § 26.604, ‘‘FFD
program requirements for facilities that
satisfy the § 26.603(c) criterion,’’ if the

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licensee or other entity performs a sitespecific analysis to demonstrate that the
facility and its operation satisfy the
criterion in § 53.860(a)(2).
Should the NRC consider replacing its
proposed § 26.603(c) criterion
referencing § 53.860(a)(2) with an
alternative requirement that if the
commercial nuclear plant is of the class
described in § 53.800, ‘‘Facility
licensees for self-reliant-mitigation
facilities,’’ and either § 53.800(a)(1) or
(2) is satisfied, then drug and alcohol
testing would not be required? This
proposal would align the § 26.603(c)
criterion with that proposed in the NRClicensed operator regulatory framework
of part 53. Please provide your
considerations and rationale for your
recommendation.
Should the NRC also consider making
a conforming change to the proposed
§ 73.120 criterion used for the AA
program? Please provide your
considerations and rationale for your
recommendation.
Part 26—Technology-Inclusive
Approaches to Fatigue Management
Requirements Applicable to Unit
Outages
In establishing the outage minimum
days off requirement of § 26.205(d)(4),
the NRC’s objective was to ensure that
individuals performing the duties
described in § 26.4(a)(1) through (a)(4)
have sufficient periodic long-duration
breaks to prevent cumulative fatigue
from degrading their ability to safely
and competently perform their duties.
In addition to the science of fatigue
management, the NRC considered
several factors in establishing the
existing requirements. These additional
factors were practical and safety
considerations associated with the
management of refueling outages for
large LWRs, including the following: (1)
the typical duration and frequency of
outages; (2) the availability of contract
personnel to perform the work; (3) the
risk presented by the outage work while
the reactor is shut down; and (4) the
controls applied to the work that may
limit the potential for latent errors to
challenge reactor safety when the
reactor is returned to power. The details
of such considerations may differ for
new reactor technologies or designs.
Such considerations may not be relevant
for some reactor designs (e.g., reactors
capable of on-line refueling) and there
may be additional, more pertinent
factors to consider for other designs.
The NRC is seeking stakeholder input
on whether alternative fatigue
management requirements applicable to
outages should be adopted to support
technology-inclusive approaches that

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would be appropriate to support the
licensing and regulation of future
commercial nuclear plants. Please
provide your considerations and
rationale for your recommendation.
Part 26—Draft Regulatory Guidance
Approach for Fatigue Management
In support of this proposed rule, the
NRC has issued DG–5078, ‘‘Fatigue
Management for Nuclear Power Plant
Personnel at Commercial Nuclear Plants
Licensed Under 10 CFR part 53.’’ This
DG describes methods the NRC staff
considers acceptable for addressing
certain aspects of FFD programs at
commercial nuclear facilities licensed
under part53.
The NRC staff also intends to
eventually transition this draft guide
into an update to RG 5.73, ‘‘Fatigue
Management for Nuclear Power Plant
Personnel,’’ or the development of a
new RG. At this point, NRC staff is
considering four options for future RG
development:
• Option 1: Amend the existing RG.
The NRC may develop an updated
version of RG 5.73 that continues to
endorse (with clarifications, additions,
and exceptions) the guidance contained
in NEI 06–11, ‘‘Managing Personnel
Fatigue at Nuclear Power Reactor Sites,’’
Revision 1, and incorporates the topics
discussed within DG–5078 as new NRC
staff positions in section C of RG 5.73.
• Option 2: Issue a new RG specific
to part 53 licensees. The NRC may
develop an entirely new RG applicable
specifically to facilities licensed under
part 53. This new RG would capture the
guidance contained in DG–5078 and
incorporate existing guidance (e.g.,
selected guidance in RG 5.73 and NEI
06–11) that is considered to be
technology inclusive in nature. The
existing guidance (i.e., RG 5.73) would
remain in place as the guidance for
facilities licensed under parts 50 and 52.
• Option 3: Review and potentially
endorse new or revised industrydeveloped guidance. The NRC may
engage with the industry regarding a
potential update to industry guidance
document NEI 06–11 or the
development of new, separate industrydeveloped guidance specific to facilities
licensed under part 53. The NRC would
then review the new or revised
industry-developed guidance within the
NRC’s RG process, which includes
opportunities for public participation.
New or revised industry-developed
guidance could incorporate DG–5078 or
propose alternatives for the NRC to
consider.
• Option 4: Develop a comprehensive
revision of the existing RG. The NRC
may develop a more comprehensive

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revision of RG 5.73 that would
explicitly detail all NRC positions
reflected in the existing RG (including
those endorsed positions currently
contained in NEI 06–11, Revision 1),
along with the guidance of DG–5078.
Such a revision would thereby be a
‘‘stand-alone’’ document, without
reference to or explicit endorsement of
separate, industry-developed guidance.
The NRC is seeking stakeholder input
regarding which of the four options
listed above would be optimal (or
whether there are other options that the
NRC should consider). Please provide
your considerations and rationale for
your recommendation.
Part 53—Overall Organization
Part 53 is structured as one framework
with subparts providing technical,
licensing, and administrative
requirements for the various stages of
the life cycle of a commercial nuclear
plant. The organization of part 53 in this
manner puts a complete set of
requirements for each stage of the life
cycle in a separate subpart with
additional subparts for licensing and
administrative requirements.
The NRC is seeking comment on the
proposed organization of the
requirements in part 53 and possible
improvements to how specific
requirements (e.g., examples of which
specific sections) could be consolidated
or otherwise reorganized to make the
rule clearer or more concise.
There are numerous references in
proposed part 53 to other NRC
regulations. Examples of such references
include those in proposed § 53.610 to
NRC regulations related to radiation
protection (part 20), FFD (part 26),
physical security (part 73), and MC&A
(10 CFR part 74, ‘‘Material Control and
Accounting of Special Nuclear
Material’’) for facilities receiving
byproduct or SNMs.
The NRC is seeking comment on
whether such references to other
regulations in various sections in the
proposed part 53 provide benefits to
applicants and licensees, or to other
stakeholders seeking to understand the
regulatory framework under part 53, or
whether such references could be
removed to reduce the length of part 53.
Part 53, Subpart B—Comprehensive
Risk Metrics
The NRC is proposing to require the
use of comprehensive risk metrics and
associated risk performance objectives
as one of several performance standards
in part 53. Comprehensive risk metrics
could include a risk metric or set of risk
metrics that approximate the total
overall risk from the facility to the

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extent practicable. Associated risk
performance objectives are
preestablished values indicative of the
comprehensive risk metrics that are
used during risk-informed decisionmaking to gauge plant safety.
Specifically, comprehensive risk metrics
and associated risk performance
objectives would provide one element of
the safety criteria for LBEs other than
DBAs in the proposed § 53.220.
Comprehensive risk metrics, in the form
of the IEFR and the ILCFR, and
associated risk performance objectives,
in the form of the QHOs of 5×10¥7 per
year and 2×10¥6 per year, respectively,
were similarly used in the LMP
methodology to ensure that other
evaluation criteria were conservatively
defined and as a tool for focusing
attention on matters important to
managing the risks posed by nuclear
power plants. The use of such
comprehensive risk metrics and
associated risk performance objectives
in an integrated risk-informed decisionmaking process is similar to that used in
RG 1.174, ‘‘An Approach for Using
Probabilistic Risk Assessment in RiskInformed Decisions on Plant-Specific
Changes to the Licensing Basis,’’
Revision 3.
The NRC is seeking comment on the
use of comprehensive risk metrics and
associated risk performance objectives
in part 53 as one of several performance
standards. The IEFR and ILCFR and the
QHOs represent comprehensive risk
metrics and associated risk performance
objectives that the NRC has used for
decades in a variety of capacities. What
other performance standards could be
used to address the comprehensive risks
posed by proposed commercial nuclear
plants? Please provide your
considerations and rationale for your
recommendation.
If an applicant proposes a novel
approach to comprehensive plant risk
and the NRC approves the approach,
should the resulting NRC-approved
comprehensive plant risk metrics and
associated risk performance objectives
be codified or otherwise memorialized
over time and, if so, how?
Part 53, Subpart B—Defense in Depth
Proposed § 53.250 would establish
requirements based on the longstanding
NRC philosophy of providing defense in
depth to address uncertainties
concerning the design, operation, and
performance of commercial nuclear
plants during LBEs.
The NRC is seeking comment on the
inclusion of the proposed requirements
to assess and provide defense in depth.
The NRC is also seeking comment on
whether to include specific provisions

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in § 53.250 and subpart B to more
explicitly address the possible role of
inherent characteristics of some SSCs in
preventing or mitigating unplanned
events. The proposed § 53.250 is
worded to preclude relying on a single
engineered design feature to address the
range of LBEs other than DBAs, which
could possibly allow crediting inherent
characteristics without further lines of
defense. How could possible inherent
characteristics of SSCs be considered in
the proposed requirements in § 53.250
or in any alternative requirements for
defense in depth provided in response
to this item? Please provide your
considerations and rationale for your
recommendation.
Part 53, Subpart C—Probabilistic Risk
Assessment
Current consensus PRA standards
provide processes for appropriately
defining the scope of a PRA and
determining applicability of supporting
requirements to suit the specific needs
of a given applicant under proposed
part 53. In addition to assessing other
aspects of PRA acceptability such as
PRA peer reviews, NRC determinations
of the acceptability of such PRAs would
assess the appropriateness of the
applicant-defined scope as part of
determining the applicability of a
consensus PRA standard supporting an
application. This approach is consistent
with the current state of practice and
offers appropriate flexibility for PRAs to
be developed and assessed based on the
application they are used to support,
which includes consideration of how
PRA results and insights are relied
upon, together with factors such as
safety margin, simplicity of design, and
treatment of uncertainty.
The NRC is seeking comment on what
additional guidance, if any, is needed
regarding PRA acceptability for Part 53
applicants and licensees.
Part 53, Subparts C and D—Earthquake
Engineering
Proposed § 53.480 would establish
requirements related to seismic design
considerations. This proposed section is
intended to provide a clear connection
between siting activities and seismic
design activities and to support various
approaches to presenting seismic
hazards and addressing those hazards in
designs. The proposed requirements are
intended to provide sufficient flexibility
to allow approaches like those currently
in parts 50 and 100 or approaches that
might be endorsed by the NRC in the
future that could incorporate more risk
insights from PRAs.
The NRC is seeking comment on
whether the proposed requirements for

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earthquake engineering provide
appropriate flexibility in addressing
seismic risks while also ensuring that
the regulations continue to adequately
address seismic hazards. Please provide
your considerations and rationale for
your recommendation.
Part 53, Subpart E—Construction and
Manufacturing
1. Proposed § 53.610(b)(1)(iii) would
require procedures that describe how
construction will be controlled so as not
to impact other features important to the
design (e.g., dewatering, slope stability,
backfill, compaction, and seepage).
The NRC is seeking comment on
whether such specific requirements are
useful or whether these requirements
could be met through other
requirements proposed in part 53 or
already present in other relevant
regulations (e.g., quality assurance
requirements in appendix B to part 50).
Part 53, Subparts E and H—
Manufacturing Licenses
1. The proposed requirements
governing manufacturing are set forth in
subpart E, and the proposed
requirements governing the licensing
processes are contained in subpart H.
Some of the proposed requirements,
including provisions related to the
loading of unirradiated fuel into a
manufactured reactor, are intended to
cover a factory-fabrication model that
has been suggested for some microreactor designs. However, as written, the
proposed provisions are not limited to
any size or type of reactor.
The NRC is seeking comment on
whether the proposed regulations are
sufficient to govern various scenarios for
the possible manufacturing and
deployment of manufactured reactors.
If a comment indicates that the
proposed regulations are not sufficient,
please describe the reasons why,
including, if applicable, any plausible
scenario for which the commenter
believes the proposed regulations are
not sufficient.
2. The proposed regulations in
subpart H allow holders of or applicants
for a COL to reference an ML but do not
include such a provision for the holder
of or applicant for a CP or OL. This
proposed change from the current
relationship between subparts in part 52
and the part 50 licensing process was
made to simplify the provisions in the
proposed part 53 for licensing and
deploying manufactured reactors.
The NRC seeks comment on whether
part 53 should include provisions for an
applicant for or a holder of a CP or an
OL to reference an ML and, if so, how
this should be done.

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3. Proposed § 53.1295 states that the
holder of an ML could not begin
manufacture of a manufactured reactor
less than 6 months before the expiration
of the license. This limitation is similar
to the current restriction in § 52.177,
which states that the manufacture of a
reactor cannot begin less than 3 years
before the expiration of the license. The
restriction was revised from 3 years in
part 52 to 6 months in the proposed part
53 in recognition of the likely use of
MLs for a factory-fabrication model for
micro-reactors.
The NRC seeks comment on whether
it is necessary or appropriate to revise
the 3-year restriction in part 52 on when
manufacturing activities could begin in
relation to license expiration and, if so,
what that restriction should be.
4. Proposed § 53.1288 provides the
finality provisions for MLs and
includes, as does existing § 52.171,
limitations on the NRC’s imposition of
new requirements on either the design
or the requirements for the manufacture
of a manufactured reactor. No MLs have
been issued under part 52 and there is
no practical experience with the
proposed finality sections. While the
implications of the finality provisions
related to the design of a manufactured
reactor can reasonably be inferred from
experience with DCs and COLs, there is
no experience or available guidance
regarding finality for ‘‘requirements for
the manufacture of the manufactured
reactor.’’
The NRC is seeking comment on the
proposed finality provisions for MLs
and specifically if and how finality for
manufacturing processes might be
requested and used.
5. The NRC is seeking comment on
the proposed regulations for the loading
of fresh (unirradiated) fuel into a
manufactured reactor for subsequent
transport to a site for which the
Commission has issued a COL that
authorizes construction and operation of
a commercial nuclear plant using the
manufactured reactor. The proposed
regulation includes provisions for
loading of fuel into manufactured
reactors at a manufacturing facility prior
to transporting the fueled reactor to its
deployment site, as suggested by some
stakeholders. The NRC has historically
viewed reactor operation as including
fuel load, and existing NRC regulations
reflect this view. While the Act
authorizes the NRC to issue licenses to
manufacture production or utilization
facilities, it does not contain specific
provisions on fueling or operating
facilities licensed under an ML, and
existing ML regulations under part 52
do not include provisions for fuel load.

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The proposed rule addresses this
matter by allowing an applicant to
combine an ML with a part 70 license,
which would authorize possession of a
manufactured reactor in which the
licensee has loaded unirradiated fuel
provided at least two independent
criticality prevention mechanisms are in
place, each of which is sufficient to
prevent criticality assuming optimum
neutron moderation and neutron
reflection conditions. This requirement
would limit the possibility of creating
fission products and allow the control of
SNM, so that the loading of the fuel into
a manufactured reactor could be
governed primarily via a part 70 license
and associated regulations (including
those in subpart H of part 70).
A specific topic on which the NRC is
seeking comment is on the potential
benefits of and issues with including the
requirements of subpart H of part 70
within the proposed regulations for
loading fuel into manufactured reactors
at the manufacturing facility. For
example, should the NRC include a
threshold for including the
requirements of subpart H of part 70
and, if so, what factors and decision
criteria should be considered in such a
threshold? If a comment indicates that
the proposed regulations are not
sufficient, please describe the reasons
why, including the plausible scenarios
for which the proposed regulations
would not work or could be made to
work better.
6. Section 170, ‘‘Indemnification and
Limitation of Liability,’’ of the Act states
that each license under section 103 shall
have as a condition of the license a
requirement that the licensee have and
maintain financial protection of such
type and in such amounts as the NRC
shall require.
The NRC is seeking comment on
whether the proposed regulations
should include amounts of required
financial protections for MLs for fueled
manufactured reactors, and, if so, what
would be appropriate amounts of
required financial protection.
7. Some stakeholders have suggested
that a fueled manufactured reactor with
appropriate protections against
criticality should not be categorized as
a utilization facility under NRC
regulations or Section 11cc. of the Act.
The NRC is seeking comment on
possible approaches where the NRC
could find that a fueled manufactured
reactor would not be a utilization
facility, the basis for such a finding, and
the potential benefits of and potential
issues with such a finding.
8. Proposed requirement
§ 53.620(d)(2)(i) would require a
security program, including a physical

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security plan, for any ML authorizing
possession of a manufactured reactor
into which fuel has been loaded at the
manufacturing facility. Currently,
requirements in § 73.67(c)(1) only
require that a physical security plan be
submitted for those licensees who
possess, use, transport, or deliver to a
carrier for transport SNM of moderate
strategic significance, or 10 kg or more
of SNM of low strategic significance.
The NRC is seeking comment on
whether the proposed requirement: (1)
should be specific to the facility type
(i.e., manufacturing facility) or be
specific to the category of material being
used at the facility; (2) should apply to
all manufacturing plants, including
those at which licensees may only
possess SNM of low strategic
significance (i.e., category III), or only
those facilities for which an applicant
must submit a physical security plan
per § 73.67(c)(1); or (3) should include
more specific requirements on the
supplemental security measures that
may be needed for licensees possessing
SNM of moderate strategic significance
(i.e., category II)?
9. Proposed requirement
§ 53.620(d)(2)(i) would require a
cybersecurity program. The proposed
general cybersecurity performance
requirements would be to provide
reasonable assurance that a cyberattack
could not adversely impact the
functions performed by digital assets
used by the licensee for implementing
the physical security, radiation
monitoring, and criticality
requirements.
The NRC is seeking comment on the
following: (1) to what extent
stakeholders envision physical security
controls, radiation monitoring, and
criticality controls at a manufacturing
facility being digital; (2) to what extent
should the ML holder be required to
protect digital computer and
communications systems that impact
safety and security functions from a
cyberattack at a manufacturing facility
authorized to load fuel; and (3) whether
the rule provides sufficient clarity on
the cybersecurity measures needed for
license issuance or if additional detail
should be included either in the rule or
in guidance?
10. Proposed requirement
§ 53.620(d)(2)(i)(B) would require that
the physical security program be
designed to prevent unintended and
uncontrolled criticality events. This
would include criticality events that are
initiated maliciously.
The NRC is seeking comment on
whether the ML holder should be
required to design its security program
to protect against radiological sabotage

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(i.e., an unintended criticality event
leading to unacceptable radiological
consequences), in addition to theft and
diversion. For example, should the NRC
establish security requirements to
prevent an adversary, including an
insider, from tampering with the reactor
at a manufacturing facility or during
transport in such a way as to cause an
inadvertent criticality event? If so,
should the NRC consider factors such as
the category of fuel and the number of
reactors at a factory that can
simultaneously be loaded with fuel in
establishing the security requirements?
11. Proposed requirement
§ 53.620(d)(2)(i) would require an ML
holder to meet the performance
objectives in § 73.67. Requirements
§ 73.67(e) and § 73.67(g) include
provisions for security of category II and
category III quantities of SNM,
respectively, during transportation.
The NRC is seeking comment on the
extent to which the ML should require
ASMs (i.e., security measures above
those required by § 73.67(e) and
§ 73.67(g)) for transportation of a fueled
reactor to its place of operation. What
should those measures be?
12. Proposed requirement
§ 53.620(d)(2)(i) would require an ML
holder to meet the performance
objectives of § 73.67. For licensees
utilizing a category II quantity of SNM,
the requirement in § 73.67(d)(4) would
have the ML holder conduct a screening
to confirm the identity of an individual
prior to granting unescorted access to
the controlled access area where the
material is used or stored. The purpose
of this requirement is to both confirm
the identity of the individual and
support a determination that the
individual is trustworthy and reliable.
The NRC is seeking comment on
whether the ML requirements should
include ASMs (i.e., measures beyond
those required by § 73.67(d)(4)) in order
to provide reasonable assurance of
identity confirmation and
trustworthiness and reliability.
13. The NRC is seeking comment on
whether provisions regulating the
testing of fueled manufactured reactors
in the manufacturing facility should be
included in part 53 and, if so, what
would be practical for the holder of an
ML while also providing adequate
protection of public health and safety.
One possibility could be COLs that
would be issued to the holders of an ML
to cover low power (e.g., <5% rated
thermal power) nuclear physics testing
of fueled manufactured reactors within
the manufacturing facility prior to the
manufactured reactors being transported
to and incorporated into a commercial
nuclear plant for the purpose of energy

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production. The NRC recognizes
configuration changes are needed to
perform nuclear physics testing and is
seeking comment on what requirements
should apply to the manufactured
reactors and the manufacturing facility
during such testing (e.g., limiting power
levels). If a comment indicates that the
regulations should address limited
operations at manufacturing facilities,
please describe the likely scenarios that
would need to be addressed and suggest
what would be appropriate
requirements for such scenarios.
While an ML holder could
accomplish nuclear physics testing by
applying for a COL under the proposed
subpart H of part 53, stakeholders have
indicated that many of the requirements
would likely be unnecessary, given the
reduced risk profile posed by such
activities. Therefore, the NRC is seeking
comment on what requirements in
subpart H of part 53 should apply to
applicants for a COL who would
perform testing of fueled manufactured
reactors at the manufacturing plant.
Examples of proposed requirements that
might be relaxed or modified for
applications for low power testing at
manufacturing plants include those
related to selection of LBEs to reflect
limited inventory of radionuclides and
decay heat, aircraft impact assessments,
and earthquake engineering.
Additionally, the NRC is seeking
comment on whether several other
requirements in part 53 could be
modified for applications for a low
power testing COL at a manufacturing
facility. For example, the NRC is seeking
comment on how portions of the ML
facility used to support testing should
fall within the requirements for
construction activities under § 53.610;
whether §§ 53.710 and 53.715 (SSC
configuration control) must be
implemented to ensure portions of the
ML facility relied on to limit potential
radiological consequences from LBEs
are available to perform their safety
functions; and whether the
requirements of § 53.730 could be
modified to reflect the conditions of low
power physics testing. If a comment
indicates that some design and analysis
requirements and related application
requirements in subpart H of the
proposed part 53 are not needed for the
testing of fueled manufactured reactors,
please provide a rationale supporting
your comment and, if applicable, what
alternate requirements would be
appropriate.
Moreover, the licensing mechanism
for the facility could present unique
challenges. One option could be to issue
a low power testing COL for each fueled
manufactured reactor to be tested. This

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would comport with the agency’s
practice of issuing one license per
reactor but could prove prohibitive from
a cost standpoint and may provide very
little safety benefit if all manufactured
reactors are the same. Alternatively, one
low power testing COL could be issued
for the portions of the ML facility used
to test the fueled manufactured reactors
and allow multiple fueled manufactured
reactors to be completed and tested over
the course of the ML. Under this
approach, any ITAAC related to testing
of the fueled manufactured reactors
would need to be closed after they were
manufactured but prior to testing, and
the NRC would issue a notice of
intended operation and provide the
public an opportunity to request a
hearing on whether each fueled
manufactured reactor as constructed
complies, or on completion will
comply, with the acceptance criteria of
the license. The NRC is seeking
comment on the potential benefits and
issues with having a COL for each
fueled manufactured reactor to be tested
versus having a COL cover the testing of
multiple fueled manufactured reactors.
If a comment indicates a preference for
a particular approach, please provide a
rationale supporting the comment and
describe the specific scenarios that the
regulations need to address.
Part 53, Subpart F—Staffing and
Generally Licensed Reactor Operators
Under the Act Sections 106 and 107,
the NRC is proposing to group
commercial reactors into classes upon
the basis of the similarity of operating
and technical characteristics of the
facilities, and then to prescribe uniform
conditions for licensing individuals as
operators of any of the various classes;
determine the qualifications of such
individuals; and, for certain classes of
commercial reactors, issue general
licenses (i.e., licenses for which no
application is needed) to such
individuals allowing the individuals to
operate the commercial reactor.
1. Categories of Individuals Who May
Manipulate Facility Controls: The NRC
is proposing requirements that would
allow the manipulation of the controls
of certain facilities by GLROs in lieu of
specifically licensed reactor operators
and senior reactor operators. Reactor
operators and senior reactor operators
are the only categories of individuals
currently allowed to be licensed to
manipulate the controls of utilization
facilities under part 55.
The NRC is interested in public
perspectives on this proposed addition
of the GLRO category, particularly in
light of new reactor technologies and
concepts of operations.

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2. Criteria for GLRO Staffing: The
NRC is proposing criteria under which
facilities would be staffed by GLROs in
lieu of specifically licensed reactor
operators and senior reactor operators.
These criteria establish a new class of
self-reliant-mitigation facilities, as
defined in part 53, for which distinct
GLRO licensing and staffing
requirements would apply.
The NRC is soliciting public feedback
regarding whether these proposed
criteria are appropriate and what, if any,
alternative criteria should be
considered. Please provide your
considerations and rationale for your
answer.
3. Medical Requirements for GLROs:
Based on the proposed criteria that a
self-reliant-mitigation facility, as
defined in part 53, must meet, the NRC
is proposing not to subject GLROs to
requirements for medical fitness and
medical examination. This is in contrast
with the proposed requirements
associated with specifically licensed
reactor operators and senior reactor
operators, as well as the existing
requirements for reactor operators and
senior reactor operators under part 55.
The NRC is soliciting public feedback
regarding whether GLROs should be
subject to medical fitness and/or
medical examination requirements like
reactor operators and senior reactor
operators. Please provide your
considerations and rationale for your
answer.
4. Onshift Engineering Expertise: The
NRC is proposing to require that
engineering expertise be accounted for
within facility staffing plans. This
proposed requirement would be in lieu
of the traditional position of the Shift
Technical Advisor. The NRC is further
proposing that individuals providing
such engineering expertise would need,
among other things, to possess either a
qualifying 4-year degree or licensure as
a Professional Engineer.
The NRC is interested in feedback
from the public regarding the
appropriateness of this requirement,
including any alternatives that should
be considered. Please provide your
considerations and rationale for your
answer.
5. Use of Simulation Facilities as HFE
Testbeds: The NRC is proposing to
establish regulations pertaining to the
use of simulation facilities within the
context of the licensing programs both
for specifically licensed reactor
operators and senior reactor operators as
well as for GLROs. However, these
regulations, as currently proposed, do
not address the use of simulation
facilities within the context of serving as
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assessments. Rather, the NRC currently
envisions that the use of simulation
facilities as HFE testbeds is more
appropriately addressed via guidance
documents.
The NRC is soliciting public feedback
regarding whether simulation facility
requirements should also address the
use of simulation facilities as HFE
testbeds. Please provide your
considerations and rationale for your
answer.
Part 53, Subpart F—Emergency
Preparedness and Security Programs
1. The proposed framework for part
53 would incorporate the changes to
NRC regulations from the final
rulemaking on ‘‘Emergency
Preparedness for Small Modular
Reactors and Other New Technologies’’
(the EP for SMR/ONT rule) by including
references to § 50.160, ‘‘Emergency
preparedness for small modular
reactors, non-light-water reactors, and
non-power production or utilization
facilities,’’ and by making conforming
changes within § 50.160. The proposed
framework for part 53 would also
introduce a graded approach to physical
protection requirements that includes
the criterion in § 53.860(a)(2)(i) to
establish a class of licensees that would
not be required to protect against the
design-basis threat (DBT) of radiological
sabotage. The NRC is soliciting public
comment relating to these topics, which
could include ways that graded
approaches for both emergency
preparedness and security programs
might be assessed and considered
during the licensing process.
The NRC is seeking comment on the
sufficiency and clarity of requirements
in proposed part 53 related to the
assessments needed to support graded
emergency planning and security. If a
comment indicates that there is an issue
with the sufficiency or clarity of the
proposed regulations, please describe
the reasons why, including, if
applicable, any scenario for which the
proposed regulations are not sufficient
and possible ways to clarify the
requirements. The NRC is specifically
seeking comment on possible challenges
arising from the interactions between
the proposed regulations and related
assessments for grading the
requirements for emergency planning
and security.
2. The NRC is preparing various
guidance documents to support this
rulemaking and other ongoing or
recently completed rulemakings related
to emergency preparedness and
security. DG–5076, ‘‘Guidance for
Technology-Inclusive Requirements for
Physical Protection of Licensed

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Activities at Commercial Nuclear
Plants,’’ has been issued along with this
proposed rulemaking and public
comments are requested via this notice
on that draft guidance. The NRC is also
planning to issue a draft revision of RG
1.242, ‘‘Performance-Based Emergency
Preparedness for Small Modular
Reactors, Non-Light-Water Reactors, and
Non-Power Production or Utilization
Facilities,’’ for public comment. The
planned revision to RG 1.242 would add
guidance for part 53 applicants and
licensees.
In the staff requirements
memorandum to SECY–23–0021, the
Commission directed the NRC staff to
address the consideration of securityrelated events for an advanced reactor
that addresses security through design
and engineered safety features when it
harmonizes this rulemaking with the EP
for SMR/ONT rule. In the EP for SMR/
ONT rule, the NRC established an
alternative performance-based and riskinformed approach for emergency
planning, including determining the
need for and size of an emergency
planning zone (EPZ) to support
predetermined, prompt protective
actions. The NRC has incorporated the
relevant rule language from the EP for
SMR/ONT rule into this proposed rule
and is seeking stakeholder feedback as
to whether additional rule language
changes or additional guidance would
be beneficial.
In light of the Commission direction
and the above considerations, the NRC
is assessing how best to address the
treatment of security-related events in
emergency planning, including in the
determination of EPZ size, for reactors
licensed under part 53. Part 53 is
introducing an alternative approach to
meeting security regulations that should
be taken into consideration under
§ 50.160. Stakeholders are encouraged to
take a holistic view of the various
activities and opportunities to provide
comments on this rulemaking and
related guidance supporting this
rulemaking (e.g., DG–5076 on physical
protection requirements, future
revisions to RG 1.242). In developing
comments, the NRC urges stakeholders
to consider various scenarios that might
arise when implementing graded
approaches for security and emergency
planning for various reactor designs.
Scenarios could include the following:
• the potential consequences from
security events up to and including the
DBT of radiological sabotage are
bounded by unlikely and very unlikely
event sequences such that security
events do not need separate analyses in
the EPZ size determination;

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• the potential consequences from
security events up to and including the
DBT are not bounded by unlikely and
very unlikely event sequences but could
otherwise support a reduced EPZ size
consistent with considerations
discussed in RG 1.242 and NUREG–
0396, ‘‘Planning Basis for the
Development of State and Local
Government Radiological Emergency
Response Plans in Support of Light
Water Nuclear Power Plants’’; or
• the potential consequences from
security events up to and including the
DBT are not bounded by unlikely and
very unlikely event sequences and
warrant consideration of increasing the
size of the EPZ.
The NRC is interested in comments
on the need for additional rule language
or guidance to address graded
approaches for emergency planning and
security programs under the scenarios
described above for part 53 applicants
and licensees. Please address within the
comments any technical, policy, or legal
issues that are associated with your
suggestions.

lotter on DSK11XQN23PROD with PROPOSALS2

Part 53, Subpart F—Integrity
Assessment Program Requirements
Decades of operating experience with
LWRs suggests that phenomena such as
environmentally assisted fatigue and
chemical interactions could impact
certain SSCs during the life of a
commercial nuclear plant. Under the
existing regulatory framework,
historically, some of these phenomena
were not addressed during early
licensing reviews but were identified
and addressed later when significant
safety issues arose (e.g., see numerous
generic letters, bulletins, orders, and
development and implementation of
vessel integrity and materials reliability
programs) or a licensee voluntarily
pursued renewal of an OL under part
54. The NRC is proposing to include a
new set of programmatic requirements
for an Integrity Assessment Program that
would ensure these phenomena are
addressed early in the life of a
commercial nuclear plant licensed
under part 53. The requirements would
be provided in § 53.870.
The NRC is seeking comment on
whether the proposed requirements
under the Integrity Assessment Program
appropriately complement design
requirements to address concerns
regarding aging, cyclic or transient load
limits, and degradation mechanisms
related to chemical interactions,
operating temperatures, effects of
irradiation, and other environmental
factors. In addition, the NRC is
interested in views on whether, and if

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so how, degradation mechanisms are or
could be addressed in other programs.
Part 53, Subpart G—Decommissioning
1. On March 3, 2022, the NRC
published the proposed rule entitled
‘‘Regulatory Improvements for
Production and Utilization Facilities
Transitioning to Decommissioning’’ (87
FR 12254). This rulemaking would
amend the NRC’s current regulations to
provide an appropriate regulatory
framework for nuclear power reactors
transitioning from operations to
decommissioning. The rulemaking
would address lessons learned from
licensees that have completed or are
currently in the decommissioning
process. The NRC staff sent a draft final
rule to the Commission for its
consideration on January 31, 2024, in
SECY–24–0011, ‘‘Final Rule: Regulatory
Improvements for Production and
Utilization Facilities Transitioning to
Decommissioning (3150–AJ59; NRC–
2015–0070).’’
What aspects of this draft final rule,
if any, should be incorporated in a part
53 final rule and why?
2. Proposed § 53.1060(b) in subpart G
would require that, ‘‘No later than 30
days after the Commission publishes
notice in the Federal Register under
§ 53.1452(a), the licensee must submit a
report containing a certification that
financial assurance for
decommissioning is being provided in
an amount specified in the licensee’s
most recent updated certification,
including a copy of the financial
instrument obtained to satisfy
§ 53.1040.’’ This is similar to the current
requirement in § 50.75(e)(3) for part 52
COL holders. The NRC is seeking
comment on whether commercial
nuclear plant COL holders under part 53
should have the same requirement as
COL holders under part 52 to
demonstrate that they have financial
assurance in place no later than 30 days
after the Commission issues the notice
of intended operation under § 53.1452.
Please provide your considerations and
rationale for your answer.
Part 53, Subpart H—Licenses To
Construct and Operate Commercial
Nuclear Plants of Identical Design at
Multiple Sites
In addition to including provisions in
part 53, subpart H, for referencing ESPs,
standard design approvals, and design
certifications in applications for
commercial nuclear plants, the
proposed § 53.1470 provides optional
requirements related to the submittal
and NRC review of CP, OL, and COL
applications to construct and operate
commercial nuclear plants of identical

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design at multiple sites, similar to
requirements found in appendix N in
both 10 CFR parts 50 and 52. This
section would set out the particular
requirements and provisions applicable
to situations in which applications for
CPs and subsequent OLs, or COLs,
under this part, are filed by one or more
applicants for licenses to construct and
operate nuclear power reactors of
identical design (‘‘common design’’) to
be located at multiple sites. Hearings for
applications filed under appendix N in
both parts 50 and 52 are governed by
subpart D of part 2, as would be the case
for future part 53 applications under
proposed § 53.1470.
Under the proposed requirements in
this section, each application is to be
treated as a separate application, with
the exception of the common design,
and so would require separate
applications, separate determinations of
sufficiency for docketing, separate
notices of docketing, and so forth.
Proposed § 53.1470 would also require
that each application list all the
applications that are to be treated
together to ensure that the NRC is
clearly informed of the intentions of all
applicants. Ordinarily, the NRC would
publish in the Federal Register a
separate notification of docketing for
each application, so that delays in the
docketing of one application would not
delay the docketing and subsequent
technical review of other applications.
However, if circumstances allow (e.g.,
sufficiency review for multiple
applications are completed
simultaneously), the NRC could publish
a single notice of docketing for multiple
applications.
With regard to how the NRC would
fulfill its obligations under the National
Environmental Policy Act of 1969, as
amended, the NRC staff would prepare
a separate environmental document for
each application, but the NRC could
conduct joint scoping on environmental
issues related to the common design. If
the applications reference a standard
design certification or the use of a
manufactured reactor, then the
environmental document would need to
incorporate by reference the
environmental assessment (EA)
prepared for either the design
certification or the ML, as applicable. In
addition, § 53.1470 would require the
ACRS to report on each of the
applications, as would be required by
provisions in subpart H of part 53. Each
ACRS report would be limited to the
safety matters which are not relevant to
the common design. In addition, the
ACRS would need to issue a report on
the safety of the common design—
except for those matters relevant to the

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safety of a referenced design
certification or manufactured reactor.
Given this synopsis of how the
requirements in proposed § 53.1470
would be implemented as currently
written, the NRC is seeking comment on
whether there are opportunities to allow
added flexibility for applicants under
these provisions. This could include
consideration of whether applications
for which the ‘‘common design’’ is not
completely identical could be evaluated
under this provision and, if so, what the
process would be for determining the
appropriateness of a common review. In
addition, the NRC is interested in
feedback about the pros and cons of
requiring that applications under these
proposed provisions be submitted at the
same time versus allowing them to be
submitted on a staggered basis.
Part 53, Subparts H and I—Probabilistic
Risk Assessment Information
Proposed § 53.1239(a)(18) in subpart
H and the related references to this
proposed requirement for the holders of
OLs and COLs would require a
description of the PRA required by
§ 53.450(a), and its results to be
included in FSARs. However, guidance
documents may further clarify the
division of PRA-related information
needed to be in the FSAR, in other
possible licensing basis documents, and
controlled as plant records subject to
inspections and audits. For example, a
possible approach could be to include a
summary of the PRA results in the
FSAR and control that information
under § 53.1545 and create a separate
document related to the broader PRA
analyses and related processes as a
program document under § 53.1560. The
program document would provide more
detail than the summaries in the FSAR
but still be a much-condensed source of
information in comparison to the
documentation of the PRA. This
possible approach would reflect the role
of the PRA in the licensing process
under part 53 and in maintaining
margins to the safety and evaluation
criteria in subparts B and C but may
allow a more appropriate evaluation
process to address the particulars and
complexities of the PRA-related
documents.
The NRC is seeking comment on the
appropriate placement of PRA-related
information among various licensing
basis documents and plant records. In
addition to the placement of PRArelated information, the NRC is seeking
comment on the appropriate control of
that information and on the routine
submittal of updates to the NRC. Please
provide your considerations and
rationale for your answer.

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Part 53, Subparts H and I—Changes to
Manufacturing Licenses
Proposed § 53.1530 would not allow
the holder of an ML or the holder of a
COL using a manufactured reactor to
make changes to the design of the
manufactured reactor without
requesting a license amendment from
the NRC. The proposed requirements do
not include a specific mention of the
manufacturing processes for which the
NRC could possibly provide finality
under proposed § 53.1288.
The NRC is seeking comment on the
appropriate change control provisions
for MLs, including whether criteria
could be developed to determine when
a license amendment request would not
be required and whether those criteria
should address changes in
manufacturing processes as well as
changes in the design. Please provide
your considerations and rationale for
your recommendation.
Financial Qualifications
Utility new reactor applicants are
exempt under § 50.33(f) from financial
qualification reviews because they are
generically presumed to be financially
qualified for operations. In contrast,
merchant power plant new reactor
applicants are required under
§ 50.33(f)(2) to submit information that
demonstrates they possess or have
reasonable assurance of obtaining the
funds necessary to cover estimated
construction and operating costs for the
period of the license. A ‘‘merchant
power plant new reactor applicant’’ is a
non-rate-regulated entity (e.g., a
nonutility) that engages in the business
of production, manufacturing,
generating, buying, aggregating,
marketing, or brokering electricity for
sale at wholesale or for retail sale to the
public. Over the past decade, the agency
has heard some concerns about the
challenges that merchant power plant
applicants face in demonstrating
compliance with the current financial
qualification requirements.
Does this standard continue to pose
challenges for merchant power plant
applicants? If so, please provide a
detailed explanation of these challenges.
Should part 53 have the same
financial qualification requirements as
parts 50 and 52? Why or why not?
Are there categories of merchant new
reactor applicants for which a part 70
‘‘appears to be financially qualified’’
standard would be more appropriate? 15
If so, please explain what types of
applicants should be able to use the part
70 financial qualification standard and
15 Section

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what distinguishes these applicants
from ones that should not be able to use
this standard.
If a part 70 financial qualification
standard were to apply to a category of
merchant new reactor applicants,
should it also apply to pre-construction
license transfer applications for these
reactors? Why or why not?
Is there another standard the agency
should consider for financial
qualification of merchant new reactor
applicants? Commenters are encouraged
to provide specific suggestions and the
basis for those suggestions.
Part 73, Section 73.100—Physical
Security
The proposed § 73.100 would identify
the proposed performance-based
physical security requirements with
which future commercial power reactor
applicants or licensees’ physical
protection programs would need to
demonstrate compliance, without
prescribing the specific methods that
must be used to satisfy them. Applicants
and licensees would have increased
flexibility regarding the modern
technologies and methods that they
could use. Implementing guidance in
DG–5076 (proposed RG 5.97),
‘‘Guidance for Technology Inclusive
Requirements for Physical Protection of
Licensed Activities at Commercial
Nuclear Plants,’’ would be available to
assist applicants and licensees. For
example, DG–5076 provides detailed
guidance, including performance
standard recommendations, on the
probability of detection and alternative
sources of power for exterior intrusion
detection systems (subsection 4.1.1.1.A),
interior intrusion detection (subsection
4.1.1.1.B), intrusion assessment
(subsection 4.1.1.2.A), security
response/neutralization subsection
(4.1.1.4.A), security communication
(subsection 4.1.1.3.A), and security
delay (subsection 4.1.1.4.C).
Does the NRC’s proposed approach in
§ 73.100 provide a sufficient level of
detail to be readily understood and
easily applied to the licensing and
oversight of new and advanced power
reactors, or should the NRC consider
moving some objective and measurable
security performance standard
recommendations from the draft
implementing guidance in DG–5076
into proposed § 73.100? If so, which
objective and measurable security
performance standard recommendations
should be moved from DG–5076 to
§ 73.100? Please provide the basis for
your response.

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Part 73, Section 73.110—Cybersecurity
The proposed § 73.110 would require
licensees to demonstrate protection
against cyberattacks in a manner that is
commensurate with the potential
consequences from those attacks,
without prescribing the specific
methods that must be used to
demonstrate protection. Under proposed
§ 73.110(a), licensees would need to
ensure that digital computer and
communications systems are adequately
protected against a potential cyberattack
that would, for example, result in
adverse impacts to the physical security
digital assets used by the licensee to
prevent unauthorized removal of
material per § 53.860(a). Protecting
against such a potential cyberattack
would involve requiring cybersecurity
for SNM at a commercial nuclear reactor
licensed under part 53. Applicants and
licensees would have increased
flexibility regarding the modern
technologies and methods that they
could use for protecting against such a
potential cyberattack. Detailed
implementing guidance in DG–5075
(proposed RG 5.96), ‘‘Establishing
Cybersecurity Programs for Commercial
Nuclear Plants licensed under 10 CFR
part 53,’’ would be available to assist
applicants and licensees. For example,
DG–5075 provides guidance on the
implementation of security by design
features (e.g., facility design) for
negating the potential consequences
from such a potential cyberattack.
If a cyberattack were to compromise
the availability, integrity, or
confidentiality of data or systems
associated with security systems/
measures for the protection of SNM at
a commercial nuclear reactor licensed
under part 53, do the potential
consequences warrant requiring
cybersecurity for such material? Please
provide the basis for your response
including a detailed explanation of
challenges, if any, posed by requiring
cybersecurity for SNM at a commercial
nuclear reactor licensed under part 53.

lotter on DSK11XQN23PROD with PROPOSALS2

Recent Legislation
On July 9, 2024, the President signed
into law the Accelerating Deployment of
Versatile, Advanced Nuclear for Clean
Energy Act of 2024, also referred to as
the ADVANCE Act. Section 203,
‘‘Licensing Considerations Relating to
Use of Nuclear Energy for Nonelectric
Applications,’’ and Section 208,
‘‘Regulatory Requirements for MicroReactors,’’ of the ADVANCE Act
specifically mention the technologyinclusive regulatory framework to be
established under section 103(a)(4) of
NEIMA as a potential vehicle to be

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considered for the report to Congress
required under section 203 and a
potential vehicle to implement
strategies and guidance for the licensing
and regulation of micro-reactors
required under section 208. This
proposed rulemaking is, in part, how
the NRC is implementing section
103(a)(4) of NEIMA.
The NRC is seeking comment on how
part 53 could be revised to better enable
its potential use to implement the
ADVANCE Act. Specifically, Section
208 of the ADVANCE Act requires the
NRC to develop and implement ‘‘riskinformed and performance-based
strategies and guidance’’ in several areas
for the licensing and regulation of
micro-reactors, including with respect
to ‘‘licensing mobile deployment.’’ The
ADVANCE Act requires the NRC to
consider ‘‘the unique characteristics of
micro-reactors,’’ including physical size,
design simplicity, and source term;
opportunities to incorporate specific
improvements related to streamlining
the review process; and other policy and
licensing issues. With regard to
implementation, the ADVANCE Act
provides the NRC with three options.
The NRC may implement the developed
strategies and guidance, as appropriate,
via (1) the existing regulatory
framework, (2) the Part 53 rulemaking,
or (3) a pending or new rulemaking.
Given the language included in Section
208, the NRC is seeking comment on
how part 53 could be revised to better
address the ADVANCE Act’s
requirements related to strategies and
guidance for micro-reactors.
VII. Section-by-Section Analysis
The following paragraphs describe the
specific changes proposed by this
rulemaking.
§ 1.43 Office of Nuclear Reactor
Regulation
This proposed rule would revise
§ 1.43(a)(2) to extend the authority of
the Office of Nuclear Reactor Regulation
to regulate source, byproduct, and SNM
at facilities licensed under part 53.
§ 2.1 Scope
This proposed rule would revise
§ 2.1(e) to apply to standard design
approvals under part 53.
§ 2.4 Definitions
This proposed rule would revise § 2.4
to update the definition of ‘‘Contested
proceeding’’ to include NRC
enforcement actions against applicants
for a standard DC under part 53. It
would also update the definition of
‘‘Facility’’ to encompass utilization
facilities as defined in § 53.020 (there

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are no production facilities under part
53).
§ 2.100

Scope of Subpart

This proposed rule would revise
§ 2.100 to extend the scope of subpart A
to licenses and standard design
approvals issued under §§ 53.1200
through 53.1221.
§ 2.101

Filing of Application

This proposed rule would revise
§ 2.101 to be applicable to part 53
applicants in addition to part 50 and 52
applicants by adding references to part
53 in paragraphs (a)(3)(i), (a)(5), and
(a)(9).
§ 2.104

Notice of Hearing

This proposed rule would extend the
hearing notice requirement in § 2.104(a)
to applications concerning facilities
covered under part 53. Footnote 1 to
§ 2.104 would be revised in a
corresponding manner.
§ 2.105

Notice of Proposed Action

This proposed rule would revise
§ 2.105 to extend the requirement in
§ 2.104 to publish a notice of intended
operation or a notice of proposed action,
as applicable, to part 53 applicants in
addition to part 50 and 52 applicants by
adding corresponding references to part
53 in paragraphs (a), (a)(4), (a)(10),
(a)(12), (a)(13), and (b)(3).
§ 2.106

Notice of Issuance

This proposed rule would revise
§ 2.106 to extend the issuance notice
requirement to applications concerning
facilities covered under part 53 through
updated references in paragraphs (a)(2)
and (3), and (b)(2).
§ 2.109 Effect of Timely Renewal
Application
This proposed rule would revise
§ 2.109 to add references to part 53 in
paragraphs (b), (c), and (d) regarding the
timing of license renewal applications.
§ 2.110 Filing and Administrative
Action on Submittals for Standard
Design Approval or Early Review of Site
Suitability Issues
This proposed rule would revise
§ 2.110 to include references to part 53
in paragraphs (a)(1) and (b).
§ 2.202

Orders

This proposed rule would revise
§ 2.202(e) to add references to part 53
regarding the requirements to be
followed for orders involving the
modification of a license, COL, ESP,
standard DC rule, standard design
approval, or ML.

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§ 2.309 Hearing Requests, Petitions To
Intervene, Requirements for Standing,
and Contentions
This proposed rule would revise
§ 2.309 to include references to part 53
in paragraphs (a), (f)(1)(i), (f)(1)(vi) and
(vii), (g), (h)(2), (i)(2), and (j) regarding
a request for hearing under § 53.1452.
§ 2.310 Selection of Hearing
Procedures
This proposed rule would revise
§ 2.310 by revising paragraph (a), the
introductory text for paragraph (h), and
paragraphs (i) and (j) to incorporate
references to part 53 regarding hearing
procedures.
§ 2.329 Prehearing Conference
This proposed rule would revise
§ 2.329(a) to extend the timing
requirements for prehearing conferences
involving CPs and licenses under part
53.

§ 2.402 Separate Hearings on Separate
Issues; Consolidation of Proceedings
This proposed rule would revise
§ 2.402(a) to apply provisions regarding
separate hearings and the consolidation
of proceedings to part 53 applicants.
§ 2.403 Notice of Proposed Action on
Applications for Operating Licenses
Pursuant To Appendix N of 10 CFR Part
50
This proposed rule would revise
§ 2.403 to require the Commission to
publish a notification of proposed
action in the Federal Register after
applications under part 53 are docketed.
§ 2.404 Hearings on Applications for
Operating Licenses Pursuant to
Appendix N of 10 CFR Part 50

§ 2.339 Expedited Decision-Making
Procedure
This proposed rule would revise
§ 2.339(d) to include references to part
53 regarding expedited decision-making
procedures.

This proposed rule would revise
§ 2.404 to apply to applications for an
OL under part 53.

§ 2.340 Initial Decision in Certain
Contested Proceedings; Immediate
Effectiveness of Initial Decisions;
Issuance of Authorizations, Permits and
Licenses
This proposed rule would revise
§ 2.340 regarding initial decisions of a
presiding officer in certain contested
proceedings, the effective date of those
decisions, and the issuance of
authorizations, permits, and licenses, by
incorporating references to part 53 in
paragraphs (b), (c), (d), (f), (i), and (j).

This proposed rule would revise
§ 2.405 to be applicable to CPs, fullpower OLs, and COLs under part 53.

§ 2.341 Review of Decisions and
Actions of a Presiding Officer
This proposed rule would revise
§ 2.341(a)(1) to include an updated
reference to part 53 regarding the
allowance of a period of interim
operation.

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the hearing notice requirement to
applications concerning facilities
covered under part 53.

§ 2.400 Scope of Subpart
This proposed rule would revise
§ 2.400 to extend the scope of subpart D
of part 2 to include part 53 applicants
for licenses to construct or operate
nuclear power reactors of identical
design at multiple sites.
§ 2.401 Notice of Hearing on
Construction Permit or Combined
License Applications Pursuant to
Appendix N of 10 CFR Parts 50, 52, or
53
This proposed rule would revise the
section heading and § 2.401 to extend

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§ 2.405 Initial Decisions in
Consolidated Hearings

§ 2.406 Finality of Decisions on
Separate Issues
This proposed rule would revise
§ 2.406 to be applicable to proceedings
conducted pursuant to part 53.
§ 2.500

Scope of Subpart

This proposed rule would revise
§ 2.500 to extend the provisions of
subpart E of part 2 to include
applications for a license to
manufacture nuclear power reactors
under part 53.
§ 2.501 Notice of Hearing on
Application Under Subpart F of 10 CFR
Part 52 or 53 for a License To
Manufacture Nuclear Power Reactors
This proposed rule would revise the
section heading and § 2.501(a) by
extending its provisions to applications
for a license to manufacture nuclear
power reactors under part 53.
§ 2.643 Acceptance and Docketing of
Application for Limited Work
Authorization
This proposed rule would revise
§ 2.643(b) regarding the acceptance and
docketing of an application for a CP for
a utilization facility of the type specified
in part 53.

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§ 2.645 Notice of Hearing
This proposed rule would revise
§ 2.645(a) to incorporate a reference to
part 53.
§ 2.649 Partial Decisions on Limited
Work Authorization
This proposed rule would revise
§ 2.649 to extend its provisions to LWAs
issued under part 53.
§ 2.800 Scope and Applicability
This proposed rule would revise
§ 2.800 by revising paragraphs (c) and
(d) to incorporate references to part 53
regarding the scope and applicability of
the rulemaking procedures contained in
this subpart.
§ 2.801 Initiation of Rulemaking
This proposed rule would revise
§ 2.801 to include a reference to part 53.
§ 2.813 Written Communications
This proposed rule would revise
§ 2.813(a) to apply general requirements
for correspondence with the
Commission to communications
concerning part 53, in addition to parts
50, 52, and 100.
§ 2.1103 Scope of Subpart K
This proposed rule would revise the
first sentence of § 2.1103 to extend the
provisions of subpart K of part 2 to
licenses under part 53 to expand the
spent fuel capacity at the site of a
civilian nuclear power plant.
§ 2.1202 Authority and Role of NRC
Staff
This proposed rule would amend
§ 2.1202 by revising paragraphs (a)(1)
through (3), and (a)(6) to include
references to part 53.
§ 2.1301 Public Notice of Receipt of a
License Transfer Application
This proposed rule would revise
§ 2.1301(b) to include a corresponding
reference to license transfers under part
53 in addition to parts 50 and 52.
§ 2.1403 Authority and Role of the
NRC Staff
This proposed role would update
§ 2.1403 to specify that ‘‘significant
hazards considerations’’ has the same
meaning as defined in part 53.
§ 2.1500 Purpose and Scope
This proposed rule would revise
§ 2.1500 to extend the scope of subpart
O of part 2 to DC rulemaking hearings
under part 53.
§ 2.1502 Commission Decision To
Hold Legislative Hearing
This proposed rule would revise
§ 2.1502, paragraphs (a) and (b)(1) to

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incorporate references to part 53
regarding the Commission’s decision to
hold a DC rulemaking.
§ 10.1 Purpose
This proposed rule would revise
§ 10.1(a)(3) to include a reference to part
53.
§ 10.2 Scope
This proposed rule would revise
§ 10.2(b) to extend the scope of subpart
A to applicants and holders of licenses,
certificates, and standard design
approvals under part 53 in addition to
part 52.
§ 11.7 Definitions
This proposed rule would revise
§ 11.7 such that terms defined in part 53
have the same meaning when used in
part 11.
§ 19.2 Scope
This proposed rule would revise
§ 19.2(a) to include references to part 53.
§ 19.3 Definitions
This proposed rule would revise the
definitions of ‘‘License’’ and ‘‘Regulated
entities’’ in § 19.3 to incorporate
references to part 53.
§ 19.11 Posting of Notices to Workers
This proposed rule would amend
§ 19.11 by revising paragraphs (a), (b),
and (e)(1) to apply to applicants and
holders of licenses, permits, standard
design approvals, and standard DCs
under part 53 in addition to part 52.
§ 19.14 Presence of Representatives of
Licensees and Regulated Entities, and
Workers During Inspections
This proposed rule would revise
§ 19.14(a) to apply to applicants and
holders of a license, standard design
approval, ESP, or standard DC under
part 53 in addition to part 52.

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§ 19.20 Employee Protection
This proposed rule would revise
§ 19.20 to include a reference to
protected activities under part 53.
§ 20.1002 Scope
This proposed rule would revise the
first sentence of 10 CFR part 20,
‘‘Standards for Protection Against
Radiation,’’ § 20.1002 to extend the
scope of part 20 to apply to persons
licensed by the Commission to receive,
use, transfer, or dispose of byproduct,
source, or SNM or to operate a
production or utilization facility under
part 53.
§ 20.1003 Definitions
This proposed rule would revise
§ 20.1003 to update the definition of

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‘‘License’’ to include those issued under
part 53.

§ 20.2201 Reports of Theft or Loss of
Licensed Material

§ 20.1101
Programs

This proposed rule would revise
§ 20.2201 to include references to part
53 in paragraphs (a)(2)(i), (b)(2)(i) and
(c) regarding requirements for reports of
theft or loss of licensed material.

Radiation Protection

This proposed rule would revise
§ 20.1101(d) to exclude licensees subject
to § 53.260 from its requirements.
§ 20.1401
Scope

General Provisions and

This proposed rule would revise
§ 20.1401, paragraphs (a) and (c) to
extend the scope of subpart E of part 20
to apply to the decommissioning of
facilities licensed under part 53 and the
release of part of a facility or site for
unrestricted use in accordance with
§ 53.1080.
§ 20.1403 Criteria for License
Termination Under Restricted
Conditions
This proposed rule would revise
§ 20.1403(d) to include
decommissioning plans under part 53.
§ 20.1404 Alternate Criteria for License
Termination
This proposed rule would revise
§ 20.1404(a)(4) to include a reference to
part 53 regarding alternate criteria for
license termination.
§ 20.1406 Minimization of
Contamination
This proposed rule would revise
§ 20.1406(a) to include references to
applicants for licenses other than ESPs
or MLs under part 53. It would also
revise § 20.1406(b) to include references
to standard DCs and standard design
approvals under part 53 in addition to
part 52.
§ 20.1501

General

This proposed rule would revise
§ 20.1501(b) regarding the requirement
for retention of records from surveys
describing the location and amount of
subsurface residual radioactivity at a
site to include a reference to the
retention requirements under part 53.
§ 20.1905 Exemptions to Labeling
Requirements
This proposed rule would revise
§ 20.1905(g) to apply to facilities
licensed under part 53 in addition to
parts 50 and 52 regarding exemptions to
labeling requirements.
§ 20.2004 Treatment or Disposal by
Incineration
This proposed rule would revise
§ 20.2004(b)(1) to include references to
part 53 regarding the treatment or
disposal of waste oil by incineration.

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§ 20.2202

Notification of Incidents

This proposed rule would revise
§ 20.2202(d)(1) to add references to part
53 regarding reports to the NRC
Operations Center.
§ 20.2203 Reports of Exposures,
Radiation Levels, and Concentrations of
Radioactive Material Exceeding the
Constraints or Limits
This proposed rule would revise
§ 20.2203(c) to refer to procedures under
part 53 for reporting occurrences of
exposures, radiation levels, and
concentrations of radioactive material
exceeding the constraints or limits.
§ 20.2206 Reports of Individual
Monitoring
This proposed rule would revise
§ 20.2206(a)(1) to include a reference to
part 53.
§ 21.2

Scope

This proposed rule would revise
§ 21.2, paragraphs (a), (b), and (c) to
include references to part 53 regarding
the scope and applicability of part 21
requirements.
§ 21.3

Definitions

This proposed rule, in § 21.3 would
revise the definitions of ‘‘Basic
component,’’ ‘‘Commercial grade item,’’
‘‘Critical characteristics,’’ ‘‘Dedicating
entity,’’ ‘‘Dedication,’’ ‘‘Defect,’’ and
‘‘Substantial safety hazard’’ with
references to part 53.
§ 21.21 Notification of Failure To
Comply or Existence of a Defect and Its
Evaluation
This proposed rule would revise
§ 21.21, by incorporating references to
part 53, to update the requirements for
notifying the Commission of a failure to
comply or defect in paragraphs (a)(3)
and (d)(1).
§ 21.51 Maintenance and Inspection of
Records
This proposed rule would revise
§ 21.51(a)(4) and (5) to apply to
applicants for standard DC and
applicants or holders of a standard
design approval under part 53, in
addition to part 52, regarding the
retention of records.

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Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules
§ 21.61 Failure To Notify
This proposed rule would revise
§ 21.61(b) to include references to part
53 licensees and applicants regarding
failure to provide the notice required in
§ 21.21.
§ 25.5 Definitions
This proposed rule would update the
definition of ‘‘License’’ to include those
issued under part 53.
§ 25.17 Approval for Processing
Applicants for Access Authorization
This proposed rule would revise
§ 25.17(a) to add a reference to part 53
regarding AAs for individuals who need
access to classified information in
connection with activities under part
53.

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§ 25.35 Classified Visits
This proposed rule would update
§ 25.35(a) to apply the requirements for
classified visits to licensees, certificate
holders, and applicants under part 53 in
addition to part 52.

26.605, 26.606, 26.607, 26.608, 26.609,
26.611, 26.613, 26.617, and 26.619.

§ 26.205 to incorporate references to
§§ 26.606 and 26.202(a) and (b).

§ 26.21

§ 26.207

Fitness-for-Duty Program

This proposed rule would revise
§ 26.21 to include a reference to
§ 26.3(f).
§ 26.51

Applicability

This proposed rule would revise
§ 26.51 to extend the requirements of
subpart C of part 26 to licensees and
other entities identified in § 26.3(f) that
do not implement the requirements of
subpart M of part 26, as well as
licensees and other entities that
implement the requirements of § 26.605.
§ 26.53

General Provisions

This proposed rule would revise
§ 26.53 paragraphs (e), (g), (h), and (i) to
include references to § 26.3(f).
§ 26.63

Suitable Inquiry

This proposed rule would revise
§ 26.63(d) with a reference to § 26.3(f).
§ 26.73

Applicability

§ 26.3 Scope
This proposed rule would amend
§ 26.3 by revising paragraph (d) and
adding new paragraph (f) which would
establish the phase of construction or
operation by which applicants and
licensees under part 53 would be
required to comply with subpart M of
part 26, or all of the requirements of part
26 except subparts K and M.

This proposed rule would revise
§ 26.73 to extend the requirements of
subpart D of part 26 to licensees and
other entities identified in § 26.3(f) that
do not implement the requirements of
subpart M of part 26, as well as
licensees and other entities that
implement the requirements of
§ 26.605(b).

§ 26.4 FFD Program Applicability to
Categories of Individuals
This proposed rule would revise
paragraphs (a), (b), (c), (e), (f), (g), and
(h) of § 26.4 to include references to part
53 and provisions for implementing an
FFD program under subpart M.

This proposed rule would revise
§ 26.81 to extend the requirements of
subpart E of part 26 to licensees and
other entities identified in § 26.3(f) that
do not implement the requirements of
subpart M of part 26, as well as
licensees and other entities that
implement the requirements of § 26.605.

§ 26.5 Definitions
This proposed rule would amend
§ 26.5 by adding definitions for
‘‘Biological marker,’’ ‘‘Change,’’ ‘‘Illicit
substance,’’ ‘‘Reduction in FFD program
effectiveness,’’ and ‘‘Special Nuclear
Material.’’ It would also revise
definitions of ‘‘Constructing or
construction activities,’’ ‘‘Contractor/
vendor (C/V),’’ ‘‘Other entity,’’
‘‘Questionable validity,’’ ‘‘Reviewing
official,’’ ‘‘Safety-related structures,
systems, and components (SSCs),’’
‘‘Security-related SSCs,’’ and ‘‘Unit
outage’’ within this section.
§ 26.8 Information Collection
Requirements: OMB Approval
This proposed rule would revise
§ 26.8(b) with the new information
collection requirements contained in
proposed §§ 26.202, 26.603, 26.604,

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§ 26.81

§ 26.201

Purpose and Applicability

Applicability

This proposed rule would revise
§ 26.201 to include references to the
proposed provisions in §§ 26.3(f) and
26.202, as well as revise the
applicability of requirements in subpart
I of part 26.
§ 26.202 General Provisions for
Facilities Licensed Under Part 53
This proposed rule would add new
§ 26.202, which would require
applicable licensees under part 53 to
incorporate a policy for fatigue
management into their FFD program in
accordance with the provisions of this
section.
§ 26.205

Work Hours

This proposed rule would revise
paragraphs (d)(7)(iii) and (d)(8) of

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Waivers and Exceptions

This proposed rule would revise
§ 26.207(a)(1)(ii) to include references to
§§ 26.608 and 26.202(c) and to include
provisions for implementing certain
face-to-face supervisor assessments
using electronic communications.
§ 26.211

Fatigue Assessments

This proposed rule would revise
§ 26.211, paragraphs (a)(1), (a)(3), and
(b) to incorporate references to
§§ 26.202(c), 26.607(b), 26.608, and
26.619 and to include provisions for
implementing certain face-to-face
assessments using electronic
communications.
Subpart M—Fitness for Duty Programs
for Facilities Licensed Under Part 53
This proposed rule would add new
Subpart M of part 26 containing
§§ 26.601, 26.603, 26.604 through
26.611, 26.613, 26.615, 26.617, and
26.619, which adds an optional
technology-inclusive, risk-informed,
and performance-based approach for the
application of drug and alcohol testing
and fatigue management requirements
for facilities licensed under part 53.
§ 26.601

Applicability

This proposed rule would add
§ 26.601, which would allow a licensee
or other entity in § 26.3(f) to establish an
FFD program in accordance with the
requirements of subpart M of part 26.
§ 26.603

General Provisions

This proposed rule would add
§ 26.603, which would establish the
general requirements for implementing
an FFD program under subpart M of part
26.
§ 26.604 FFD Program Requirements
for Facilities That Satisfy the § 26.603(c)
Criterion
This proposed rule would add
§ 26.604, which would establish the
FFD program elements for a licensee or
other entity whose facilities and
operations demonstrate compliance
with the criterion in § 26.603(c).
§ 26.605 FFD Program Requirements
for Facilities That Do Not Implement
§ 26.604
This proposed rule would add
§ 26.605, which would establish the
FFD program elements for a licensee or
other entity that does not demonstrate
compliance with the criterion in
§ 26.603(c), or otherwise chooses to
maintain an FFD program under this
section.

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§ 26.606 Written Policies and
Procedures
This proposed rule would add
§ 26.606, which would require licensees
and other entities that implement an
FFD program under subpart M of part 26
to develop a written FFD policy
statement and provide it to all
individuals subject to the FFD program,
and to establish, implement, and
maintain written procedures addressing
the topics outlined in this section.

audits to monitor the effectiveness of
FFD program elements.

§ 26.607 Drug and Alcohol Testing
This proposed rule would add
§ 26.607, which would establish
requirements for licensees and other
entities performing drug and alcohol
testing as part of an FFD program under
subpart M of part 26.

§ 26.619 Suitability and Fitness
Determinations

§ 26.608 FFD Program Training
This proposed rule would add
§ 26.608, which would require
individuals who are subject to the FFD
program under subpart M of part 26 to
receive periodic training on FFD
policies and procedures, including their
duties and responsibilities under the
BOP.
§ 26.609 Behavioral Observation
This proposed rule would add
§ 26.609, which would establish the
requirements for a BOP under subpart M
of part 26.
§ 26.610 Sanctions
This proposed rule would add
§ 26.610, which would require licensees
and other entities implementing an FFD
program under subpart M of part 26 to
establish sanctions for FFD policy
violations.

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§ 26.611 Protection of Information
This proposed rule would add
§ 26.611, which would require licensees
and other entities implementing an FFD
program under subpart M of part 26 to
establish a system to protect personal
information against unauthorized
disclosure.
§ 26.613 Appeals Process
This proposed rule would add
§ 26.613, which would require licensees
and other entities that implement an
FFD program under subpart M of part 26
to establish procedures for an individual
to appeal a policy violation
determination.
§ 26.615 Audits
This proposed rule would add
§ 26.615, which would establish
provisions for licensees and other
entities that implement an FFD program
under subpart M of part 26 to conduct

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§ 26.617

Recordkeeping and Reporting

This proposed rule would add
§ 26.617, which would require licensees
or other entities implementing an FFD
program under subpart M of part 26 to
retain records pertaining to the
administration of the program and to
make reports in accordance with the
requirements of this section.

This proposed rule would add
§ 26.619, which would require licensees
and other entities that implement FFD
programs to develop, implement, and
maintain procedures to assess whether
individuals are fit to perform the duties
that make them subject to the FFD
program.
§ 26.709

Applicability

This proposed rule would designate
the current paragraph as new paragraph
(a), and it would be revised to reference
paragraphs (a) through (d) of § 26.3. It
would also add paragraph (b) to
§ 26.709, which would extend the
requirements of subpart N of part 26 to
licensees and other entities identified in
§ 26.3(f) that do not implement the
requirements of subpart M of part 26, as
well as licensees and other entities that
implement the requirements of
§ 26.605(b).
§ 26.711

General Provisions

This proposed rule would revise
§ 26.711(c) and (d) to incorporate a
reference to § 26.3(f).
§ 26.825

Criminal Penalties

This proposed rule would revise
§ 26.825(b) to include a reference to the
proposed § 26.601.
§ 30.4

Definitions

This proposed rule would revise the
definition for ‘‘Utilization facility’’ in
§ 30.4 to include utilization facilities
defined in the regulations under part 53
in addition to part 50.
§ 30.50

Reporting Requirements

This proposed rule would revise
§ 30.50(c)(3) to include references to
part 53 in addition to part 50.
§ 40.60

Reporting Requirements

This proposed rule would revise
§ 40.60(c)(3) to include references to
part 53 in addition to part 50 regarding
reporting requirements.

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§ 50.47

Emergency Plans

This proposed rule would revise
§ 50.47(a)(1) and (e) with appropriate
references to part 53.
§ 50.54

Conditions of Licenses

This proposed rule would revise
§ 50.54(q)(2), (q)(4), and (gg)(1) with
appropriate references to part 53.
§ 50.160 Emergency Preparedness for
Small Modular Reactors, Non-LightWater Reactors, and Non-Power
Production or Utilization Facilities
This proposed rule would revise
§ 50.160(b)(3) and (c)(2) with the
appropriate references to part 53.
Appendix B to 10 CFR Part 50—Quality
Assurance Criteria for Nuclear Power
Plants and Fuel Reprocessing Plants
This proposed rule would revise
appendix B to part 50 by revising the
introduction and specific criteria to
incorporate the appropriate references
and terminology for part 53.
§ 51.20 Criteria for and Identification
of Licensing and Regulatory Actions
Requiring Environmental Impact
Statements
This proposed rule would revise
§ 51.20(b)(1) and (2) to require an EIS
prior to the issuance of a CP, LWA, or
ESP under part 53, or the issuance to
renewal of a full power or design
capacity license to operate a nuclear
power reactor, testing facility, or fuel
reprocessing plant under part 53.
§ 51.22 Criterion for Categorical
Exclusion; Identification of Licensing
and Regulatory Actions Eligible for
Categorical Exclusion or Otherwise Not
Requiring Environmental Review
This proposed rule would revise
§ 51.22 to include corresponding
references to part 53 in paragraphs
(c)(3), (c)(9), (c)(12), (c)(17), (c)(22) and
(23).
§ 51.26 Requirement To Publish Notice
of Intent and Conduct Scoping Process
This proposed rule would revise
§ 51.26(d) to add a reference to part 53.
§ 51.30

Environmental Assessment

This proposed rule would revise the
introductory text to paragraph (a) and
revise paragraphs (d) and (e) of § 51.30
to incorporate the appropriate
references to part 53 regarding EAs.
§ 51.31 Determinations Based on
Environmental Assessment
This proposed rule would revise
§ 51.31(a) to include a reference to part
53.

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§ 51.32
Impact

§ 51.95 Postconstruction
Environmental Impact Statements

Finding of No Significant

This proposed rule would revise
§ 51.32(b)(1) and (3), finding there is no
significant environmental impact
associated with the issuance of standard
DCs and MLs under part 53.
§ 51.49 Environmental Report-Limited
Work Authorization
This proposed rule would revise the
introductory text of § 51.49(c) to require
applicants for an ESP under part 53
requesting a LWA to include the
environmental report required by
§ 51.50(b).
§ 51.50 Environmental Report—
Construction Permit, Early Site Permit,
or Combined License Stage
This proposed rule would revise
§ 51.50, paragraphs (a), (b)(4), and the
introductory text for paragraph (c) to
incorporate the appropriate references
to part 53.
§ 51.53 Postconstruction
Environmental Reports
This proposed rule would revise
§ 51.53(d) to include the appropriate
references to part 53 regarding a license
termination plan or decommissioning
plan and related requirements for
postconstruction environmental reports.
§ 51.54 Environmental Report—
Manufacturing License
This proposed rule would update
§ 51.54(a) to require applicants for MLs
under part 53 to submit an
environmental report with the
application.
§ 51.55 Environmental Report—
Standard Design Certification
This proposed rule would update
§ 51.55(a) to require applicants for a
standard DC under part 53 to submit an
environmental report with the
application.
§ 51.58 Environmental Report—
Number of Copies; Distribution
This proposed rule would revise
§ 51.58(b) to incorporate the appropriate
references to part 53.

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§ 51.77 Distribution of Draft
Environmental Impact Statement

§ 51.92 Supplement to the Final
Environmental Impact Statement
This proposed rule would revise
§ 51.92(b) to apply to COL applications
referencing an ESP under part 53.

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§ 51.101

Limitations on Actions

This proposed rule would revise
§ 51.101(a)(2) to include the
corresponding references to part 53
where appropriate.
§ 51.103

Record of Decision—General

This proposed rule would update
§ 51.103(a)(6) to apply to the issuance of
a LWA in connection with a CP or COL
under part 53.
§ 51.105 Public Hearings in
Proceedings for Issuance of
Construction Permits or Early Site
Permits; Limited Work Authorizations
This proposed rule would revise
§ 51.105(c)(1) to include the appropriate
reference to LWAs under part 53 for CPs
or ESPs.
§ 51.107 Public Hearings in
Proceedings for Issuance of Combined
Licenses; Limited Work Authorizations
This proposed rule would amend
§ 51.107 by revising the introductory
text for paragraphs (a) and (b) and
updating paragraph (d)(1) to include the
appropriate corresponding references to
part 53.
§ 51.108 Public Hearings on
Commission Findings That Inspections,
Tests, Analyses, and Acceptance
Criteria of Combined Licenses Are Met
This proposed rule would revise
§ 51.108 to incorporate the appropriate
references to part 53.
10 CFR part 53—Risk-Informed,
Technology-Inclusive Regulatory
Framework for Commercial Nuclear
Plants
This proposed rule would add a new
part to 10 CFR Chapter I, designated as
Part 53 including §§ 53.000 through
53.9010.
§ 53.000

This proposed rule would revise the
introductory text for § 51.77(a) to add a
reference to part 53.

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This proposed rule would revise the
introductory text for § 51.95(c) to
include a reference to part 53 regarding
the Commission’s obligations to prepare
an EIS following the renewal of an
operating or COL for a nuclear plant
under part 53.

Purpose

This proposed rule would add
§ 53.000 which provides an optional
technology-inclusive, performancebased framework for the issuance,
amendment, renewal, and termination
of licenses, permits, certifications, and
approvals for commercial nuclear plants
licensed under section 103 of the
Atomic Energy Act of 1954, as amended.

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86993

Subpart A—General Provisions
This proposed rule would add subpart
A, to establish a set of general
provisions, which apply to all
applicants and licensees under part 53.
§ 53.015

Scope

This proposed rule would add
§ 53.015, which would extend the
provisions of subpart A to all applicants
and licensees under part 53.
§ 53.020

Definitions

This proposed rule would add
§ 53.020, which would define key terms
in part 53.
§ 53.040

Written Communications

This proposed rule would add
§ 53.040, which would govern how
applicants and licensees submit written
communications to the NRC, including
applications, submissions related to the
security plans, emergency plan, and
quality assurance, certifications of
permanent cessation of operations and
permanent fuel removal, and other
submittals required under part 53.
§ 53.050

Deliberate Misconduct

This proposed rule would add
§ 53.050, which would prohibit
licensees or applicants, contractors and
subcontractors, or employees of those
entities from deliberately violating NRC
rules, regulations, or orders, or the
terms, conditions, and limitations of a
part 53 license. This proposed rule
would also prohibit deliberate
submissions of incomplete or inaccurate
information. Violations would be
subject to enforcement actions under
subpart B of part 2.
§ 53.060

Employee Protection

This proposed rule would add
§ 53.060, which would prohibit
applicants and licensees from
discriminating against employees for
engaging in the protected activities
listed in this section and provide
remedial procedures for employees who
believe they are the subjects of
discrimination.
§ 53.070 Completeness and Accuracy
of Information
This proposed rule would add
§ 53.070, which would require licensees
and applicants under part 53 to provide
complete and accurate information in
accordance with all applicable laws,
Commission regulations, and the terms
and conditions of their license. This
proposed rule would also require
licensees to notify the Commission
within two days of identifying
information with material implications

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for public health and safety or common
defense and security.
§ 53.080

Specific Exemptions

This proposed rule would add
§ 53.080, which would establish the
special circumstances under which the
Commission could grant exemptions to
part 53 licensees and the Commission’s
criteria for making such a
determination.
§ 53.090

Standards for Review

This proposed rule would add
§ 53.090 to establish the standards that
the Commission would consider when
determining whether to issue a permit
or license under part 53.
§ 53.100

Jurisdictional Limits

This proposed rule would add
§ 53.100, which would provide that
permits, licenses, standard design
approvals, and standard DCs are solely
issued for activities within the
jurisdiction of the United States.
§ 53.110

Attacks and Destructive Acts

This proposed rule would add
§ 53.110, which would exempt licensees
or applicants under part 53 from
providing design features to protect
against attacks or destructive acts
directed at the facility by United States
adversaries.
§ 53.115 Rights Related to Special
Nuclear Material
This proposed rule would add
§ 53.115, which would establish
provisions regarding the rights to SNM
under a part 53 license.
§ 53.117 License Suspension and
Rights of Recapture
This proposed rule would add
§ 53.117, which would provide that the
Commission may suspend licenses and
recapture material or control of a facility
in a state of war or national emergency
declared by Congress.
§ 53.120 Information Collection
Requirements: OMB Approval

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This proposed rule would add
§ 53.120, which would establish
requirements for information collection
requirements and Office of Budget and
Management approval.

§ 53.210 Safety Criteria for DesignBasis Accidents
This proposed rule would add
§ 53.210 to set dose values to ensure that
plants are designed to limit the public’s
radiation exposure in the event of a
DBA.
§ 53.220 Safety Criteria for LicensingBasis Events Other Than Design-Basis
Accidents
This proposed rule would add
§ 53.220 to require plants to implement
a combination of design features and
programmatic controls to control risks
to the public in the event of a LBE other
than a DBA.
§ 53.230 Safety Functions
This proposed rule would add
§ 53.230, which specifies that limiting
the release of radioactive materials from
the facility is the primary safety
function of a commercial nuclear plant,
and that additional safety functions
must be defined to support the retention
of radioactive materials during LBEs.
§ 53.240 Licensing-Basis Events
This proposed rule would add
§ 53.240 to require commercial nuclear
plants to conduct an analysis of LBEs to
confirm that design features and
programmatic controls satisfy the safety
criteria under §§ 53.210 and 53.220, or
alternatively, under § 53.470.
§ 53.250 Defense in Depth
This proposed rule would add
§ 53.250 to establish a performancebased, defense-in-depth approach to
address uncertainties about the
effectiveness and reliability of plant
SSCs, personnel, and programmatic
controls.
§ 53.260 Normal Operations
This proposed rule would add
§ 53.260, requiring holders of licenses to
operate commercial nuclear plants to
control public doses and dose rates in
unrestricted areas to meet the
requirements in part 20, during normal
plant operation.

Subpart B—Technology-Inclusive Safety
Requirements

§ 53.270 Protection of Plant Workers
This proposed rule would add
§ 53.270, requiring holders of licenses to
operate commercial nuclear plants to
control occupational doses to meet the
requirements in part 20.

This proposed rule would add subpart
B, to establish a set of technologyinclusive performance standards that
would be used throughout part 53 to
determine appropriate regulatory
controls for SSCs, human actions, and
programs.

Subpart C—Design and Analysis
Requirements
This proposed rule would add subpart
C, which requires the implementation of
certain design features and the
performance of risk assessments and
analyses to demonstrate compliance

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with the safety criteria and safety
functions in subpart B.
§ 53.400 Design Features for LicensingBasis Events
This proposed rule would add
§ 53.400, which would require design
features that satisfy the safety criteria
defined in § 53.210 and § 53.220 or
§ 53.470 and fulfill the safety functions
identified in § 53.230 during LBEs.
§ 53.410 Functional Design Criteria for
Design-Basis Accidents
This proposed rule would add
§ 53.410, which would stipulate that
functional design criteria must be
defined for each design feature required
by § 53.400 to demonstrate compliance
with the safety criteria defined in
§ 53.210 for DBAs.
§ 53.415 Protection Against External
Hazards
This proposed rule would add
§ 53.415, which would require SR SSCs
to be designed to withstand the effects
of natural phenomena and constructed
hazards while performing the intended
safety functions.
§ 53.420 Functional Design Criteria for
Licensing-Basis Events Other Than
Design-Basis Accidents
This proposed rule would add
§ 53.420, which would require
functional design criteria to be defined
for each design feature required by
§ 53.400 to demonstrate compliance
with the safety criteria defined in
§ 53.220 for LBEs other than DBAs.
§ 53.425 Design Features and
Functional Design Criteria for Normal
Operations
This proposed rule would add
§ 53.425, which would require
commercial nuclear plants to implement
design features and define functional
design criteria sufficient to demonstrate
compliance with § 53.850 and show
through functional design criteria that
design features and corresponding
programmatic controls control wastes,
as required under part 20.
§ 53.430 Design Features and
Functional Design Criteria for Protection
of Plant Workers
This proposed rule would add
§ 53.430, which would require
commercial nuclear plants to implement
design features and define functional
design criteria sufficient to demonstrate
compliance with § 53.270.
§ 53.440

Design Requirements

This proposed rule would add
§ 53.440, which would establish various

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design feature requirements, including
protection against fires and explosions,
criticality accidents, and the impact of
a large commercial aircraft.
§ 53.450

Analysis Requirements

This proposed rule would add
§ 53.450, which would require
commercial nuclear plants to perform
PRAs in combination with other
analytical methods to identify and
assess risks and determine compliance
with the safety criteria in subpart B. In
addition, § 53.450 would require
analysis of DBAs and other analyses to
assess the adequacy of protections
against fire, aircraft impact, and the
release of effluents.
§ 53.460 Safety Categorization and
Special Treatments
This proposed rule would add
§ 53.460 to address the safety
classification of SSCs and determine
appropriate special treatments.

This proposed rule would add
§ 53.470 to permit applicants and
licensees to implement more restrictive
criteria than that defined in §§ 53.220
and 53.450(e) to support operational
flexibilities.
Earthquake Engineering

This proposed rule would add
§ 53.480 to provide overall seismic
design considerations based on the
safety criteria in subpart B and siting
requirements in subpart D to ensure that
SSCs are able to withstand the effects of
earthquakes without loss of capability to
fulfill safety functions.
Subpart D—Siting Requirements
This proposed rule would add subpart
D, which would address requirements
associated with the siting of commercial
nuclear facilities under part 53,
including considerations of external
hazards and potential adverse impacts
on the surrounding population.

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§ 53.500 General Siting and Siting
Assessment
This proposed rule would add
§ 53.500, which would require a siting
assessment for each commercial nuclear
plant to ensure that design features and
programmatic controls are sufficient to
address LBEs and mitigate potential
adverse impacts of the plant on the
surrounding environs.
§ 53.510

External Hazards

This proposed rule would add
§ 53.510, which would require site-

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§ 53.520

Site Characteristics

This proposed rule would add
§ 53.520, which would require the
design and analyses conducted under
subpart C to consider how site
characteristics may contribute to LBEs.
§ 53.530 Population-Related
Considerations
This proposed rule would add
§ 53.530, which would establish
requirements related to the facility’s
exclusion area, low-population zone,
and population center distance.
§ 53.540

§ 53.470 Maintaining Analytical Safety
Margins Used To Justify Operational
Flexibilities

§ 53.480

specific assessments, including an
evaluation of geological and seismic
siting factors, to identify and
characterize the external hazard level
for a range of natural and constructed
hazards.

Siting Interfaces

This proposed rule would add
§ 53.540, which would require that
external hazards and site characteristics
must be accounted for in the design
features, programmatic controls, and
supporting analyses used to
demonstrate compliance with the safety
criteria in §§ 53.210 and 53.220.
Subpart E—Construction and
Manufacturing Requirements
This proposed rule would add subpart
E, which would establish requirements
for the construction and manufacture of
commercial nuclear plants.
§ 53.600 Construction and
Manufacturing—Scope and Purpose
This proposed rule would add
§ 53.600, which would indicate that this
subpart applies to construction and
manufacturing activities authorized by a
CP, COL, ML, or LWA issued under this
part.
§ 53.605 Reporting of Defects and
Noncompliance
This proposed rule would add
§ 53.605, which would describe the
procedures, notification requirements,
and records retention requirements that
each CP, ML, and COL is subject to with
respect to reporting of defects and
noncompliance.
§ 53.610

Construction

This proposed rule adds § 53.610 to
address the management and control of
the construction of a commercial
nuclear plant, including specific
requirements for procedures and quality
assurance, control of radioactive
materials, and post construction
inspections.

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§ 53.620 Manufacturing
This proposed rule would add
§ 53.620, which would ensure that the
holders of an ML under part 53 develop
plans, programs, and organizational
units to manage and control
manufacturing activities, and would
establish requirements for the loading of
fuel into a manufactured reactor for
subsequent transport to a commercial
nuclear plant and operation pursuant to
a COL.
Subpart F—Requirements for Operation
This proposed rule would add subpart
F, which would establish regulatory
requirements to ensure that the safety
criteria in subpart B are satisfied
whenever a commercial nuclear plant
licensed under part 53 is operational.
This includes periods of normal
operation and unplanned events.
§ 53.700 Operational Objectives
This proposed rule would add
§ 53.700, which would establish general
operational objectives to ensure that
licensees under part 53 have
implemented and maintained the SSCs
necessary to demonstrate compliance
with the safety functions identified in
subpart B for addressing normal
operations and responding to LBEs.
§ 53.710 Maintaining Capabilities and
Availability of Structures, Systems, and
Components
This proposed rule would add
§ 53.710, which would require licensees
under part 53 to demonstrate
compliance with the safety criteria in
subpart B by establishing TS for all SR
SSCs and developing documents and
procedures for all NSRSS SSCs.
§ 53.715 Maintenance, Repair, and
Inspection Programs
This proposed rule would add
§ 53.715, which would require licensees
to develop, implement, and maintain
programs to assess and manage any risks
posed by maintenance activities and to
evaluate the efficacy of performance,
condition monitoring, and maintenance
activities.
§ 53.720 Response to Seismic Events
This proposed rule would add
§ 53.720, which would establish
requirements for licensees to respond to
a seismic event during the operating
phase of the life cycle of a commercial
nuclear plant.
§ 53.725 General Staffing, Training,
Personnel Qualifications, and Human
Factors Requirements
This proposed rule would add
§ 53.725, which would provide an

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overview of the staffing, training,
personnel qualifications, and human
factors requirements established in
§§ 53.725 through 53.830 and would
provide definitions of ‘‘Automation,’’
‘‘Auxiliary operator,’’ ‘‘Controls,’’
‘‘Generally licensed reactor operator,’’
‘‘Load following,’’ ‘‘Operator,’’
‘‘Performance testing,’’ ‘‘Reference
plant,’’ ‘‘Self-reliant mitigation facility,’’
‘‘Senior operator,’’ ‘‘Simulation
facility,’’ and ‘‘Systems approach to
training.’’ Proposed §§ 53.725 through
53.830 would apply to applicants for or
holders of OLs or COLs under part 53.
§ 53.726

Communications

This proposed rule would add
§ 53.726, which would contain
communications requirements
applicable to sections §§ 53.725 through
53.830. It also contains requirements to
notify the Commission within 30 days
should a specifically licensed operator
or senior operator be reassigned,
terminated, or suffer permanent
disability or illness.
§ 53.728 Completeness and Accuracy
of Information
This proposed rule would add
§ 53.728, which would require
submitted information to be complete
and accurate in all material respects.

This proposed rule would add
§ 53.730, which would establish
technical requirements for applicants or
holders of OLs or COLs within the areas
of HFE, human-system interface design,
concept of operations, functional
requirements analysis, function
allocation, operating experience,
procedures, staffing, operator training,
operator examinations, and operator
proficiency.

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Operator Licensing

This proposed rule would add
§ 53.760, which would address the
applicability of the requirements of
§§ 53.760 through 53.795 for specifically
licensed operators and senior operators.
§ 53.765

Medical Requirements

This proposed rule would add
§ 53.765, which would establish
medical requirements for specifically
licensed operators and senior operators.
§ 53.770 Incapacitation Because of
Disability or Illness
This proposed rule would add
§ 53.770, which would establish
requirements to address permanent
medical conditions for specifically
licensed operators and senior operators.
§ 53.775 Applications for Operators
and Senior Operators
This proposed rule would add
§ 53.775, which would establish the
application process and requirements
for individuals applying for specific
operator and senior operator licenses.
§ 53.780 Training, Examination, and
Proficiency Program

§ 53.730 Defining, Fulfilling, and
Maintaining the Role of Personnel in
Ensuring Safe Operations

§ 53.735

§ 53.760

General Exemptions

This proposed rule would add
§ 53.780, which would contain the
requirements associated with
specifically licensed operator and senior
operator initial training, initial
examinations, requalification training,
requalification examinations,
examination integrity, simulation
facilities, waivers, and proficiency.
§ 53.785 Conditions of Operator and
Senior Operator Licenses
This proposed rule would add
§ 53.785, which would establish
conditions for specific operator and
senior operator licenses.

This proposed rule would add
§ 53.735, which would establish general
exemptions for licensed operators.

§ 53.790 Issuance, Modification, and
Revocation of Operator and Senior
Operator Licenses

§ 53.740 Facility Licensee
Requirements—General

This proposed rule would add
§ 53.790, which would contain
requirements associated with the
issuance, modification, or revocation of
specific operator and senior operator
licenses.

This proposed rule would add
§ 53.740, which would establish staffing
requirements for interaction-dependentmitigation facilities and self-reliant
mitigation facilities.
§ 53.745 Operator License
Requirements
This proposed rule would add
§ 53.745, which would require
individuals to be licensed to perform
certain functions.

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§ 53.795 Expiration and Renewal of
Operator and Senior Operator Licenses
This proposed rule would add
§ 53.795, which would contain
requirements associated with the
expiration and renewal of specific
operator and senior operator licenses.

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§ 53.800 Facility Licensees for SelfReliant-Mitigation Facilities
This proposed rule would add
§ 53.800, which would establish the
technical criteria by which commercial
nuclear plants under part 53 are
determined to be of the self-reliant
mitigation class of facilities that would
be staffed by GLROs in lieu of
specifically licensed operators and
senior operators.
§ 53.805 Facility Licensee
Requirements Related to Generally
Licensed Reactor Operators
This proposed rule would add
§ 53.805, which would establish
requirements that apply to the facility
licensee at those facilities staffed by
GLROs.
§ 53.810 Generally Licensed Reactor
Operators
This proposed rule would add
§ 53.810, which would issue and
describe the general license for GLROs
that manipulate the controls of a selfreliant mitigation facility.
§ 53.815 Generally Licensed Reactor
Operator Training, Examination, and
Proficiency Programs
This proposed rule would add
§ 53.815, which would contain the
requirements for GLRO initial training,
initial examinations, continuing
training, requalification examinations,
examination integrity, simulation
facilities, examination waivers, and
proficiency.
§ 53.820 Cessation of Individual
Applicability
This proposed rule would add
§ 53.820, which would address the
requirements by which the general
license for GLROs would cease to be
applicable on an individual basis.
§ 53.830 Training and Qualification of
Commercial Nuclear Plant Personnel
This proposed rule would add
§ 53.830, which would address training
and qualification requirements for
supervisors, technicians, and other
appropriate operating personnel at
commercial nuclear plants.
§ 53.845

Programs

This proposed rule would add
§ 53.845, which would require licensees
under part 53 to establish programs that
include, but are not limited to, radiation
protection, emergency preparedness,
security, quality assurance, integrity
assessment, fire protection, ISI and IST,
and facility safety, to ensure that the
safety criteria and functions in subpart

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B are maintained during normal
operations and LBEs.

Subpart G—Decommissioning
Requirements

for using decommissioning trust funds
and required terms.

§ 53.850

This proposed rule would add subpart
G, to establish decommissioning
requirements for applicants for or
holders of an OL or COL under part 53.

§ 53.1050

Radiation Protection

This proposed rule would add
§ 53.850, which would require licensees
under part 53 to implement and
maintain programs and processes to
limit and monitor radioactive plant
effluents and limit the exposure of plant
personnel and the public.
§ 53.855

Emergency Preparedness

This proposed rule would add
§ 53.855, which would require licensees
under this part to have an emergency
response plan for radiological
emergencies.
§ 53.860

Security Programs

This proposed rule would add
§ 53.860, which would require licensees
under part 53 to develop, implement,
and maintain programs for physical
security, FFD, AA, cybersecurity, and
information security.
§ 53.865

Quality Assurance

This proposed rule would add
§ 53.865, which would require licensees
under part 53 to establish a quality
assurance program that includes a
written manual to ensure activities are
conducted in accordance with codes
and standards found acceptable by the
NRC.
§ 53.870 Integrity Assessment
Programs
This proposed rule would add
§ 53.870, which would require licensees
under part 53 to establish an integrity
assessment program to ensure that the
plant continues to fulfill safety criteria
and functional design criteria as it ages.
§ 53.875

Fire Protection

This proposed rule would add
§ 53.875, which would require licensees
under part 53 to establish a fire
protection plan and describe the
necessary elements that the plan must
incorporate.
§ 53.880 Inservice Inspection and
Inservice Testing
This proposed rule would add
§ 53.880, which would require licensees
under part 53 to develop and implement
a program for ISI and IST in accordance
with the requirements of this section.
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86997

§ 53.910

Procedures and Guidelines

This proposed rule would add
§ 53.910, which would require licensees
under part 53 to develop, maintain, and
implement procedures and guidelines
that address normal plant operations
and responses to unplanned events.

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§ 53.1000

Scope and Purpose

This proposed rule would add
§ 53.1000, which would establish the
scope of the decommissioning
requirements for applicants and
licensees under part 53 and describe the
contents of subpart G of part 53.
§ 53.1010 Financial Assurance for
Decommissioning
This proposed rule would add
§ 53.1010, which would establish the
requirement that applicants for an OL or
COL under part 53 provide reasonable
assurance that funds will be available
for the decommissioning process. This
section would describe the requirements
associated with the required plan and
an associated decommissioning report
that ensures and documents that
adequate funding for decommissioning
will be available.
§ 53.1020 Cost Estimates for
Decommissioning
This proposed rule would add
§ 53.1020, which would require sitespecific cost estimates for
decommissioning and establish the
aspects that must be included in the
estimate.
§ 53.1030 Annual Adjustments to Cost
Estimates for Decommissioning
This proposed rule would add
§ 53.1030, which would require that
holders of an OL or COL under part 53
annually adjust their cost estimate for
decommissioning to account for
escalation in labor, energy, and waste
burial costs. This section would allow
licensees to elect either a site-specific
adjustment factor or a generic
adjustment factor.
§ 53.1040 Methods for Providing
Financial Assurance for
Decommissioning
This proposed rule would add
§ 53.1040, which would establish
suitable methods that holders of an OL
or COL under part 53 may use to
provide financial assurance for
decommissioning to the NRC.
§ 53.1045 Limitations on the Use of
Decommissioning Trust Funds
This proposed rule would add
§ 53.1045, which would establish
requirements for decommissioning trust
funds under part 53, including criteria

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NRC Oversight

This proposed rule would add
§ 53.1050, which would outline the
steps the NRC may take to ensure
adequate accumulation of
decommissioning funds.
§ 53.1060 Reporting and
Recordkeeping Requirements
This proposed rule would add
§ 53.1060, which would contain
reporting and recordkeeping
requirements related to
decommissioning for each holder of an
OL or COL under part 53. This section
would outline requirements for
documents such as: certification of
decommissioning funding,
decommissioning cost estimates and
copies of financial instruments, licensee
records of information important to safe
and effective decommissioning, postshutdown decommissioning activities
report, financial assurance reports, and
reports on the status of funding for
managing irradiated fuel.
§ 53.1070

Termination of License

This proposed rule would add
§ 53.1070, which would establish
procedures for decommissioning and
license termination applicable to
licensees under part 53 that have
determined to permanently cease
operations.
§ 53.1075 Program Requirements
During Decommissioning
This proposed rule would add
§ 53.1075, which would require
licensees under part 53 to establish and
maintain a decommissioning fire
protection program to prevent, detect,
and control fires, and ensure that the
risk of fire induced radiological hazards
are minimized through the various
stages of facility decommissioning.
§ 53.1080 Release of Part of a
Commercial Nuclear Plant or Site for
Unrestricted Use
This proposed rule would add
§ 53.1080, which would establish
licensee procedures for requesting and
NRC procedures for approving partial
release of a commercial nuclear plant or
site for unrestricted use prior to
receiving approval of a license
termination plan from the Commission
under part 53.
Subpart H—Licenses, Certifications, and
Approvals
This proposed rule would add subpart
H, which would govern the process of
applying for, amending, renewing, or

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terminating a LWA, ESP, standard
design approval, standard DC, ML, CP,
OL, or COL under part 53.
§ 53.1100 Filling of Application for
Licenses, Certifications, or Approvals;
Oath or Affirmation
This proposed rule would add
§ 53.1100, which would establish
requirements for applicants seeking a
standard design approval, standard DC,
license, or permit under part 53 to
submit an application.
§ 53.1101

Requirement for License

This proposed rule would add
§ 53.1101, which would prohibit any
use of a utilization facility except as
authorized by a license issued by the
NRC or by an exception as described in
§ 53.1120.
§ 53.1103
Licenses

Combining Applications and

filed an application for a CP or COL for
that site.

This proposed rule would add
§ 53.1115, which would require
applicants to agree in writing, prior to
receiving a license or standard design
approval under part 53, to restrict any
facilities, or any individuals with access
to plant facilities, from possessing
Restricted Data or classified National
Security Information until they have
received the appropriate authorization.

§ 53.1144 Contents of Applications for
Early Site Permits; General Information

§ 53.1118 Ineligibility of Certain
Applicants
This proposed rule would add
§ 53.1118, which would prevent
citizens, nationals, or agents of a foreign
country or corporations owned,
controlled, or dominated by a foreign
entity from applying for or obtaining a
license under part 53.
§ 53.1120 Exceptions and Exemptions
From Licensing Requirements

This proposed rule would add
§ 53.1103, which would permit
applicants under part 53 seeking
multiple licenses to submit a single
application, and the Commission to
issue a single license for activities that
would otherwise be licensed separately.

This proposed rule would add
§ 53.1120, which would establish the
activities that are exempt from licensing
requirements.

§ 53.1106

This proposed rule would add
§ 53.1121, which would allow applicant
submissions to be made publicly
available under the provisions of part 2.

Elimination of Repetition

This proposed rule would add
§ 53.1106, which would allow
applicants under part 53 to reference
information contained in previous
documents filed with the Commission
so long as those references are clear and
specific.
§ 53.1109 Contents of Applications;
General Information
This proposed rule would add
§ 53.1109, which would establish the
general content to be included in
applications made under part 53,
including but not limited to the
identifying information of the applicant
and the radiological emergency
response plans of government entities
within the plume exposure pathway
EPZ.
§ 53.1112

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§ 53.1115 Agreement Limiting Access
to Classified Information

Environmental Conditions

This proposed rule would add
§ 53.1112, which would allow the
Commission to attach conditions to CPs,
ESPs, and licenses issued under part 53
to address environmental issues during
construction, operation, or
decommissioning. These conditions will
be derived from the information
contained in the environmental report
submitted as part of the application for
a permit or license.

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§ 53.1121 Public Inspection of
Applications

§ 53.1124
Sections

Relationship Between

This proposed rule would add
§ 53.1124, which would outline the
relationship between LWAs, ESPs,
standard design approvals, standard
DCs, MLs, CPs, OLs, and COLs under
part 53.
§ 53.1130 Limited Work
Authorizations
This proposed rule would add
§ 53.1130, which would establish
requirements for requesting an LWA
and grounds for the Commission to
issue an LWA. It would also contain
details about the effect of an LWA and
the implementation of a redress plan.
§ 53.1140

Early Site Permits

This proposed rule would add
§ 53.1140, which would provide an
overview of the requirements regarding
applications for and the issuance of
ESPs under part 53.
§ 53.1143

Filing of Applications

This proposed rule would add
§ 53.1143, which would enable an
applicant under part 53 to apply for an
ESP, regardless of whether they have

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This proposed rule would add
§ 53.1144, which would require
applications for ESPs to include the
information required by § 53.1109(a)
through (d) and (j).
§ 53.1146 Contents of Applications for
Early Site Permits; Technical
Information
This proposed rule would add
§ 53.1146, which would require
applicants for ESPs to submit technical
information, including but not limited
to a Site Safety Analysis Report and
emergency plans.
§ 53.1149

Review of Applications

This proposed rule would add
§ 53.1149, which would establish
standards for review of applications for
ESPs under part 53, including
requirements for the Commission to
prepare an EIS and assess the adequacy
of protective actions in the event of a
radiological emergency. It would also
require the administrative review of
applications and hearings to follow the
procedural requirements of part 2.
§ 53.1155 Referral to the Advisory
Committee on Reactor Safeguards
This proposed rule would add
§ 53.1115, which would require the
ACRS to review SR content in the
application for an ESP under part 53.
§ 53.1158

Issuance of Early Site Permit

This proposed rule would add
§ 53.1158, which would establish the
conditions under which the
Commission may issue an ESP under
part 53, as well as the information,
terms, and conditions to be included in
the permit.
§ 53.1161 Extent of Activities
Permitted
This proposed rule would add
§ 53.1161, which would require that a
valid ESP only be used for the purpose
of site redress, unless the site is
referenced in an application for a CP or
COL under part 53.
§ 53.1164

Duration of Permit

This proposed rule would add
§ 53.1164, which would govern the
conditions under which an ESP remains
valid following the date of issuance.
§ 53.1167 Limited Work Authorization
After Issuance of Early Site Permit
This proposed rule would add
§ 53.1167, which would permit the

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holder of an ESP to request a LWA
under § 53.1130.
§ 53.1170

Transfer of Early Site Permit

This proposed rule would add
§ 53.1170, which would govern the
transfer of an ESP in accordance with
§ 53.1570.
§ 53.1173

Application for Renewal

This proposed rule would add
§ 53.1173, which would establish the
conditions and procedures for renewing
an ESP under part 53.
§ 53.1176

Criteria for Renewal

This proposed rule would add
§ 53.1176, which would establish the
criteria that the Commission may use to
grant a renewal of an ESP under part 53.
§ 53.1179

Duration of Renewal

This proposed rule would add
§ 53.1179, which would govern the
duration of a renewed ESP under part
53.
§ 53.1182
Purposes

Use of Site for Other

This proposed rule would add
§ 53.1182, which would govern
acceptable uses of the site for purposes
other than those described in the
permit.
§ 53.1188 Finality of Early Site Permit
Determinations
This proposed rule would add
§ 53.1188, which would address the
finality of ESP determinations under
part 53.
§ 53.1200

Standard Design Approvals

This proposed rule would add
§ 53.1200, which would address the
procedures for filing an application for
a standard design approval under part
53, the process of review by NRC staff,
and referral to the ACRS of standard
designs.
§ 53.1203

Filing of Applications

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This proposed rule would add
§ 53.1203, which would enable
applicants to submit a final design for
the entire facility, or major portions, to
the NRC staff for review.
§ 53.1206 Contents of Applications for
Standard Design Approvals; General
Information
This proposed rule would add
§ 53.1206, which would require
applications for a standard design
approval under part 53 to contain the
information required by § 53.1109(a)
through (c) and (j).

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§ 53.1209 Contents of Applications for
Standard Design Approvals; Technical
Information
This proposed rule would add
§ 53.1209, which would require the
inclusion of certain technical
information, including a FSAR, site
parameters, and design information,
when an applicant seeks review of
major portions of a standard design.
§ 53.1210 Contents of Applications for
Standard Design Approvals; Other
Application Content
This proposed rule would add
§ 53.1210, which would require
applications for standard design
approvals under part 53 to include a
description of the availability controls
used to satisfy the safety criteria of
§ 53.220, the program to protect
Safeguards Information against
unauthorized disclosure, evidence that
safety questions associated with SSCs
have been resolved, and a description of
how design features fulfill design
criteria.
§ 53.1212 Standards for Review of
Applications
This proposed rule would add
§ 53.1212, which would require
applications for standard design
approval to be reviewed under the
standards in parts 20, 53, and 73.
§ 53.1215 Referral to the Advisory
Committee on Reactor Safeguards
This proposed rule would add
§ 53.1215, which would require the
ACRS to report on any portions of the
application for a standard design
approval under part 53 concerning
safety.
§ 53.1218 Staff Approval of Design
This proposed rule would add
§ 53.1218, which would require the NRC
staff to make a determination on the
acceptability of the design, publish its
decision in the Federal Register, and
issue a report analyzing the design that
is available at http://nrc.gov.
Additionally, the rule would establish
the conditions under which a design
approval under part 53 remains valid.
§ 53.1221 Finality of Standard Design
Approvals; Information Requests
This proposed rule would add
§ 53.1221, which would require NRC
staff and the ACRS to rely upon an
approved design in their review of any
standard DC, ML, or individual facility
license application under part 53 that
references the standard design approval.
The proposed rule would also govern
requirements for issuing information
requests.

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§ 53.1230 Standard Design
Certifications
This proposed rule would add
§ 53.1230, which would provide an
overview of the requirements and
procedures that govern the issuance of
standard DCs under part 53.
§ 53.1233

Filing of Applications

This proposed rule would add
§ 53.1233, which would enable an
application for DC to be filed, regardless
of whether an application for a CP, COL,
or ML has been filed, provided it
complies with the filing requirements in
§ 53.040 and §§ 2.811 through 2.819.
§ 53.1236 Contents of Applications for
Standard Design Certifications; General
Information
This proposed rule would add
§ 53.1236, which would require an
application for a standard DC under part
53 to contain all of the information
required by § 53.1109(a) through (c) and
(j).
§ 53.1239 Contents of Applications for
Standard Design Certifications;
Technical Information
This proposed rule would add
§ 53.1239, which would require
applicants for a standard DC under part
53 to submit a FSAR that includes
technical design information at a level
of detail sufficient to enable the
Commission to make a safety
determination.
§ 53.1241 Contents of Applications for
Standard Design Certifications; Other
Application Content
This proposed rule would add
§ 53.1241, which would require
applications for standard DCs under
part 53 to include an environmental
report, as well as a description of the
availability controls used to satisfy the
safety criteria of § 53.220, proposed
ITAAC, the program to protect
Safeguards Information against
unauthorized disclosure, evidence that
safety questions associated with SSCs
have been resolved, and a description of
how design features fulfill design
criteria.
§ 53.1242

Review of Applications

This proposed rule would add
§ 53.1242, which would require
applications for standard DCs to be
reviewed for compliance with the
standards in parts 20, 51, 53, and 73. It
would also establish procedural
requirements for reviewing applications
and holding hearings in accordance
with subpart H of part 2.

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§ 53.1273

§ 53.1245 Referral to the Advisory
Committee on Reactor Safeguards
This proposed rule would add
§ 53.1245, which would require the
ACRS to report on any portions of the
application for a standard DC under part
53 concerning safety.
§ 53.1248 Issuance of Standard Design
Certification
This proposed rule would add
§ 53.1248, which would establish the
conditions under which the
Commission may issue a DC rule that
specifies the site parameters, design
characteristics, and any additional terms
and conditions of the DC rule.
§ 53.1251

Duration of Certification

This proposed rule would add
§ 53.1251, which would set the
conditions under which a standard DC
remains valid.
§ 53.1254

Application for Renewal

This proposed rule would add
§ 53.1254, which would establish the
conditions and procedures for renewing
a standard DC under part 53.
§ 53.1257

Criteria for Renewal

This proposed rule would add
§ 53.1257, which would enable the
Commission to issue a rule granting the
renewal of a standard DC under part 53,
impose additional requirements, and
grant amendment requests.
§ 53.1260

Duration of Renewal

This proposed rule would add
§ 53.1260, which would provide that a
renewal of a standard DC under part 53
is valid for not less than 10 years, nor
more than 15 years.

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§ 53.1263 Finality of Standard Design
Certifications
This proposed rule would add
§ 53.1263, which would establish
limited conditions under which the
Commission may initiate a rulemaking
to modify, rescind, or impose new
requirements on a standard DC rule
under part 53. It would also address
requests for an exemption from
elements of the certification
information, and require that applicants
for a CP, COL, or ML that references a
DC rule make information normally
contained in engineering documents
available for audit.
§ 53.1270

Manufacturing Licenses

This proposed rule would add
§ 53.1270, which would provide an
overview of the requirements and
procedures for applying for and issuing
an ML under part 53.

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Filing of Applications

This proposed rule would add
§ 53.1273, which would establish the
requirements to apply for an ML under
part 53.
§ 53.1276 Contents of Applications for
Manufacturing Licenses; General
Information
This proposed rule would add
§ 53.1276, which would require
applicants for an ML under part 53 to
include the information contained in
§ 53.1109(a) through (e) and (j).
§ 53.1279 Contents of Applications for
Manufacturing Licenses; Technical
Information
This proposed rule would add
§ 53.1279, which would require an
applicant for an ML under part 53 to
include certain technical information in
a FSAR, including but not limited to
information about site parameters,
design information, manufacturing
information, and information related to
the potential fueling and ultimate
deployment of a completed
manufactured reactor.
§ 53.1282 Contents of Applications for
Manufacturing Licenses; Other
Application Content
This proposed rule would add
§ 53.1282, which would require
applicants for an ML under part 53 to
include in their application the
proposed ITAAC, an environmental
report, a description of the program to
protect Safeguards Information against
unauthorized disclosure, and a
description of how design features
fulfill design criteria. It would also
include content requirements for the
ITAAC and environmental reports in
applications that reference a standard
DC.
§ 53.1285

Review of Applications

This proposed rule would add
§ 53.1285, which would require
applications for MLs under part 53 to be
reviewed for compliance with
applicable standards and establish
procedural requirements for reviewing
applicants and holding hearings in
accordance with part 2.
§ 53.1286 Referral to the Advisory
Committee on Reactor Safeguards
This proposed rule would add
§ 53.1286, which would require the
ACRS to report on any portions of the
application for an ML under part 53
concerning safety.

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§ 53.1287 Issuance of Manufacturing
Licenses
This proposed rule would add
§ 53.1287, which would establish the
conditions under which the
Commission may issue an ML under
part 53.
§ 53.1288 Finality of Manufacturing
Licenses
This proposed rule would add
§ 53.1288, which would address the
limited circumstances in which the
Commission may modify, rescind, or
impose new requirements following the
issuance of an ML under part 53. It
would also address requests for a
departure from the specifications of the
license.
§ 53.1291 Duration of Manufacturing
Licenses
This proposed rule would add
§ 53.1291, which would govern the
expiration of an ML, which is valid for
no less than 5, nor more than 15 years
from the date of issuance.
§ 53.1293 Transfer of Manufacturing
Licenses
This proposed rule would add
§ 53.1293, which would provide that an
ML under part 53 may be transferred in
accordance with § 53.1570.
§ 53.1295 Renewal of Manufacturing
Licenses
This proposed rule would add
§ 53.1295, which would establish the
procedures for applicants to apply for
and the Commission to grant a renewal
of an ML under part 53.
§ 53.1300 Construction Permits
This proposed rule would add
§ 53.1300, which would provide an
overview of the requirements and
procedures for applicants to apply for
and the Commission to grant a CP under
part 53.
§ 53.1306 Contents of Applications for
Construction Permits; General
Information
This proposed rule would add
§ 53.1306, which would require
applicants for a CP under part 53 to
submit the general information required
by § 53.1109, as well as financial
information.
§ 53.1309 Contents of Applications for
Construction Permits; Technical
Information
This proposed rule would add
§ 53.1309, which would require
applicants for a CP under part 53 to
submit a PSAR and a description of the
program to protect Safeguards

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Information from unauthorized
disclosure.
§ 53.1312 Contents of Applications for
Construction Permits; Other Application
Content
This proposed rule would add
§ 53.1312, which would require
applicants for a CP under part 53 to
submit an environmental report and to
provide additional details in the PSAR
if the application references an ESP,
standard design approval, or standard
DC.
§ 53.1315

Review of Applications

This proposed rule would add
§ 53.1315, which would require
applications for CPs under part 53 to be
reviewed for compliance with
applicable standards and establish
procedural requirements for reviewing
applications and holding hearings in
accordance with part 2.
§ 53.1318 Finality of Referenced NRC
Approvals, Permits, and Certifications
This proposed rule would add
§ 53.1318, which would address the
finality of ESPs, standard design
approvals, and standard DCs referenced
in the CP application.
§ 53.1324 Referral to the Advisory
Committee on Reactor Safeguards

§ 53.1327 Authorization To Conduct
Limited Work Authorization Activities

§ 53.1330 Exemptions, Departures,
and Variances

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This proposed rule would add
§ 53.1330, which would govern requests
for and issuance of exemptions from the
Commission’s regulations and
exemptions, departures, and variances
from NRC approvals, permits, and
certifications.

This proposed rule would add
§ 53.1333, which would establish the
conditions under which the
Commission may issue CPs and
accompanying terms and conditions
under part 53.

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§ 53.1345 Transfer of Construction
Permits
This proposed rule would add
§ 53.1345, which would govern the
transfer of CPs under part 53.
§ 53.1348 Termination of Construction
Permits
This proposed rule would add
§ 53.1348, which would require the
holder of a permit under part 53 to
provide written certification to the
Commission within 30 days of
determining to permanently cease
construction.
§ 53.1360

Operating Licenses

This proposed rule would add
§ 53.1360, which would provide an
overview of the requirements and
procedures for applicants to apply for
and the Commission to issue an OL
under part 53.

This proposed rule would add
§ 53.1366, which would require an
application for an OL under part 53 to
include the information required by
§ 53.1109 as well as financial
information.
§ 53.1369 Contents of Applications for
Operating Licenses; Technical
Information

This proposed rule would add
§ 53.1327, which would govern
authorization to conduct LWA
activities.

Issuance of Construction

§ 53.1342 Duration of Construction
Permits
This proposed rule would add
§ 53.1342, which would establish
requirements for the expiration of a CP.

§ 53.1366 Contents of Applications for
Operating Licenses; General Information

This proposed rule would add
§ 53.1324, which would require the
ACRS to report on any portions of the
application for a CP under part 53
concerning safety.

§ 53.1333
Permits

§ 53.1336 Finality of Construction
Permits
This proposed rule would add
§ 53.1336, which would address the
finality of CPs.

This proposed rule would add
§ 53.1369, which would require an
application for an OL under part 53 to
include certain technical information in
an FSAR at a level of detail sufficient for
the Commission to reach a final
conclusion on all safety matters.
§ 53.1372 Contents of Applications for
Operating Licenses; Other Application
Content
This proposed rule would add
§ 53.1372, which would require an
application for an OL under part 53 to
include an environmental report and a
description of availability controls.
§ 53.1375

Review of Applications

This proposed rule would add
§ 53.1375, which would establish the
standards and procedures for reviewing

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87001

applications and holding hearings on
OLs under part 53.
§ 53.1381 Referral to the Advisory
Committee on Reactor Safeguards
This proposed rule would add
§ 53.1381, which would require the
ACRS to report on any portions of the
application for a CP under part 53
concerning safety.
§ 53.1384 Exemptions, Departures,
and Variances
This proposed rule would add
§ 53.1384, which would govern requests
for and the issuance of exemptions from
the Commission’s regulations and
exemptions, departures, and variances
from NRC approvals, permits, and
certifications.
§ 53.1387
Licenses

Issuance of Operating

This proposed rule would add
§ 53.1387, which would establish the
conditions under which the
Commission may issue OLs and
accompanying conditions and
limitations, including TS, under part 53.
§ 53.1390
Licenses

Backfitting of Operating

This proposed rule would add
§ 53.1390, which would prevent the
Commission from modifying, adding, or
deleting any terms or conditions of the
OL, except in accordance with
§ 53.1590.
§ 53.1396
Licenses

Duration of Operating

This proposed rule would add
§ 53.1396, which would provide that an
OL under part 53 may be valid for up
to 40 years.
§ 53.1399
License

Transfer of an Operating

This proposed rule would add
§ 53.1399, which would provide that an
OL under part 53 may be transferred
under § 53.1570.
§ 53.1402

Application for Renewal

This proposed rule would add
§ 53.1402, which would provide that an
application for a renewed OL under part
53 must be filed in accordance with
§ 53.1595.
§ 53.1405 Continuation of an
Operating License
This proposed rule would add
§ 53.1405, which would govern the
continuing obligations of the holder of
an OL under part 53 following the
permanent cessation of operations.

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§ 53.1410

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§ 53.1434 Authorization To Conduct
Limited Work Authorization Activities

Combined Licenses

This proposed rule would add
§ 53.1410, which would provide an
overview of the requirements and
procedures for applicants to apply for
and the Commission to issue a COL
under part 53.

This proposed rule would add
§ 53.1434, which would address
authorization to conduct LWA
activities.

§ 53.1413 Contents of Applications for
Combined Licenses; General
Information
This proposed rule would add
§ 53.1413, which would require an
application for a COL under part 53 to
include the information required by
§ 53.1109 as well as financial
information.
§ 53.1416 Contents of Applications for
Combined Licenses; Technical
Information
This proposed rule would add
§ 53.1416, which would require
applicants for a COL under part 53 to
submit an FSAR with a level of
technical information sufficient to reach
a final conclusion on all safety matters.
§ 53.1419 Contents of Applications for
Combined Licenses; Other Application
Content
This proposed rule would add
§ 53.1419, which would require
applicants for a COL under part 53 to
submit an environmental report, a
description of availability controls, the
ITAAC that the licensee must perform.
It would also include ITAAC
requirements for applications that
reference an ESP, standard DC, ML, or
combination thereof.
§ 53.1422

Review of Applications

This proposed rule would add
§ 53.1422, which would require
applications for COLs under part 53 to
be reviewed for compliance with
applicable standards and establish
procedural requirements for reviewing
applications and holding hearings in
accordance with part 2.
§ 53.1425 Finality of Referenced NRC
Approvals

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This proposed rule would add
§ 53.1425 which would address the
finality of ESPs, standard DC rules,
standard design approvals, or MLs
referenced in the application for a COL
under part 53.
§ 53.1431 Referral to the Advisory
Committee on Reactor Safeguards
This proposed rule would add
§ 53.1431, which would require the
ACRS to report on any portions of the
application for a COL under part 53
concerning safety.

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§ 53.1437 Exemptions, Departures,
and Variances
This proposed rule would add
§ 53.1437, which would govern the
conditions in which the Commission
may grant an exemption for one or more
of its regulations, or an exemption,
variance, or departure from a permit,
design approval, or license.
§ 53.1440
Licenses

Issuance of Combined

This proposed rule would add
§ 53.1440, which would establish the
conditions under which the
Commission may issue COLs and
accompanying conditions and
limitations, including TS, under part 53.
§ 53.1443
Licenses

Finality of Combined

This proposed rule would add
§ 53.1443, which would govern
permissible modifications or
amendments that the Commission may
make to a COL, as well as permissible
changes that a licensee may make to
facilities and procedures as described in
the FSAR.
§ 53.1449 Inspection During
Construction
This proposed rule would add
§ 53.1449, which would establish
requirements related to inspections,
tests, or analyses for the holder of a COL
under part 53.
§ 53.1452 Operation Under a
Combined License
This proposed rule would add
§ 53.1452, which would establish
requirements describing the
notifications, hearings, and findings to
be made prior to commencing facility
operations.
§ 53.1455
License

Duration of a Combined

This proposed rule would add
§ 53.1455, which would govern the
duration of a COL under part 53.
§ 53.1456
License

Transfer of a Combined

This proposed rule would add
§ 53.1456, which would permit the
transfer of a COL under part 53 in
accordance with § 53.1570.

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§ 53.1458 Application for Renewal
This proposed rule would add
§ 53.1458, which would provide that an
application for renewal of a COL must
be filed in accordance with § 53.1595.
§ 53.1461 Continuation of Combined
License
This proposed rule would add
§ 53.1461, which would govern the
continuing obligations of the holder of
a COL under part 53 following the
permanent cessation of operations.
§ 53.1470 Standardization of
Commercial Nuclear Plant Designs:
Licenses To Construct and Operate
Nuclear Power Reactors of Identical
Design at Multiple Sites
This proposed rule would add
§ 53.1470, which would govern the
requirements and procedures for filing
and issuing applications for a CP, OL, or
COL under part 53 in which the
applicant seeks approval of the same
design for multiple sites.
Subpart I—Maintaining and Revising
Licensing-Basis Information
This proposed rule would add subpart
I, which would address the maintenance
of licensing-basis information for part
53.
§ 53.1500 Licensing-Basis Information
This proposed rule would add
§ 53.1500, describing the purpose of
subpart I, which would be to provide
the requirements for the maintenance of
licensing-basis information for
commercial nuclear plants licensed
under part 53.
§ 53.1502 Specific Terms and
Conditions of Licenses
This proposed rule would add
§ 53.1502, which would outline the
specific terms and conditions for
obtaining a license under part 53.
§ 53.1505 Changes to Licensing-Basis
Information Requiring Prior NRC
Approval
This proposed rule would add
§ 53.1505, which would provide an
overview of the process for licensees to
request, and the Commission to issue,
amendments to licensing-basis
information under part 53.
§ 53.1510 Application for Amendment
of License
This proposed rule would add
§ 53.1510, which would require
licensees under part 53 to file an
application to request an amendment to
the license. Applicants must assess how
their requested changes would impact
the safety criteria and analysis

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§ 53.1545 Updating Final Safety
Analysis Reports

requirements in subpart B and C, as
applicable, whether the amendment
involves no significant hazards
consideration using the standards in
§ 53.1520 and consider potential
impacts on environmental factors.

This proposed rule would add
§ 53.1545, which would require
licensees under part 53 to regularly
update FSARs in accordance with the
requirements of this section to reflect
changes to licensing-basis information.

§ 53.1515 Public Notices; State
Consultation
This proposed rule would add
§ 53.1515, which would outline the
Commission’s procedures for issuing a
notification in the Federal Register and
consulting with the State in which the
commercial nuclear facility is located in
connection with its consideration of
applications for an amendment to an OL
or COL under part 53.
§ 53.1520

Issuance of Amendment

This proposed rule would add
§ 53.1520, which would outline criteria
for the Commission to consider in
issuing license amendments under part
53.
§ 53.1525 Revising Certification
Information Within a Design
Certification Rule
This proposed rule would add
§ 53.1525, which would address the
requirements for applicants to request,
and the Commission to grant, an
exemption to a DC rule under part 53.
§ 53.1530 Revising Design Information
Within a Manufacturing License
This proposed rule would add
§ 53.1530, which would require the
holder of an ML to request an
amendment under § 53.1510 and, as
applicable, § 53.1520 to make changes to
the design of a manufactured reactor. It
would also outline the requirements for
holders of a COL under part 53 to
request amendments for changes to the
design information of a manufactured
reactor.
§ 53.1535 Amendments During
Construction

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This proposed rule would add
§ 53.1535, which would outline the
process for licensees under part 53 to
request amendments to CPs or LWAs
during construction.

This proposed rule would add
§ 53.1540, which would provide an
overview of the regulations in subpart I
for holders of an OL or COL under part
53 to modify licensing-basis information
and definitions relevant to §§ 53.1545
through 53.1565.

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This proposed rule would add
§ 53.1550, which would require
licensees under part 53 to follow the
guidelines outlined in this section in
determining whether changes to
licensing-basis information described in
the FSAR (as updated) require them to
obtain a license amendment.
§ 53.1560 Updating Program
Documents Included in Licensing-Basis
Information
This proposed rule would add
§ 53.1560, which would require the
holders of an OL or COL under part 53
to regularly update the program
documents that they submitted in their
application for a license.
§ 53.1565 Evaluating Changes to
Programs Included in Licensing-Basis
Information
This proposed rule would add
§ 53.1565, which would enable
licensees under part 53 to make changes
to the facility, procedures, or
organization, or address changes to site
environs as described in program
documents without NRC approval if
these changes satisfy the criteria
outlined in this section.
§ 53.1570

Transfer of Licenses

This proposed rule would add
§ 53.1570, which would outline the
requirements for an application for
transfer of a license issued under part
53.
§ 53.1575

§ 53.1540 Updating Licensing-Basis
Information and Determining the Need
for NRC Approval

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§ 53.1550 Evaluating Changes to
Facility as Described in Final Safety
Analysis Reports

Termination of Licenses

This proposed rule would add
§ 53.1575, which would outline the
process for terminating an OL or COL
issued under part 53.
§ 53.1580

Information Requests

This proposed rule would add
§ 53.1580, which would address the
process and circumstances under which
the NRC may send information requests
to the various types of licensees within
part 53.

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§ 53.1585 Revocation, Suspension,
Modification of Licenses and Approvals
for Cause
This proposed rule would add
§ 53.1585, which would address
grounds for the revocation, suspension,
or modification of a license or standard
design approval issued under part 53.
§ 53.1590 Backfitting
This proposed rule would add
§ 53.1590, which would define
backfitting and establish requirements
to be met by the NRC when it takes
backfitting actions under part 53.
§ 53.1595 Renewal
This proposed rule would add
§ 53.1595, which would provide for the
renewal of a license under part 53 upon
expiration.
Subpart J—Reporting and Other
Administrative Requirements
This proposed rule would add subpart
J, to establish various reporting and
other administrative requirements for
licensees under part 53.
§ 53.1600 General Information
This proposed rule would add
§ 53.1600, which provides an overview
of the sections that would require
applicants and licensees under part 53
to provide NRC inspectors with
unfettered access to sites and facilities,
maintain records and make reports,
demonstrate compliance with financial
qualification and reporting
requirements, and maintain required
financial protection for accidents.
§ 53.1610 Unfettered Access for
Inspections
This proposed rule would add
§ 53.1610, which would require
applicants and licensees under part 53
to provide unfettered access to NRC
inspectors, including access to records,
premises, activities, and licensed
materials, in addition to providing office
space to accommodate temporary or
resident inspectors.
§ 53.1620 Maintenance of Records,
Making of Reports
This proposed rule would add
§ 53.1620, which would require part 53
licensees to maintain all records and
make reports as required by the
conditions of the license or by the
regulations in part 53.
§ 53.1630 Immediate Notification
Requirements for Operating Commercial
Nuclear Plants
This proposed rule would add
§ 53.1630, which would impose
immediate notification requirements on

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part 53 licensees following the
declaration of an Emergency Class or the
discovery of certain non-emergency
events.

financial qualifications information
outlined in this section within seventyfive days of ceasing to be an electric
utility.

(k) to include the appropriate references
to part 53.

§ 53.1640
System

§ 53.1700

This proposed rule would revise
§ 70.24(d) to include the appropriate
references to part 53.

Licensee Event Report

This proposed rule would add
§ 53.1640, which would require any
commercial plant licensee holding an
OL under part 53 to submit a Licensee
Event Report in accordance with the
specifications outlined in this section.
§ 53.1645 Reports of Radiation
Exposure to Members of the Public
The proposed rule would add
§ 53.1645, which would require annual
reports to the Commission, including
radiological reports as required by part
20, an Annual Radioactive Effluent
Release Report, and an Annual
Environmental Operating Report.
§ 53.1650 Facility Information and
Verification
The proposed rule would add
§ 53.1650, which would include a
reporting requirement for applicants
and holders of a CP or license under
part 53 to support safeguards
agreements between the United States
and the IAEA.
§ 53.1660

Financial Requirements

This proposed rule would add
§ 53.1660, which would introduce
requirements and procedures related to
financial qualifications and reporting
requirements for applicants, licensees,
and CP holders under part 53.
§ 53.1670

Financial Qualifications

This proposed rule would add
§ 53.1670, which would require an
applicant for a CP, OL, or COL under
part 53 to must demonstrate possession
or ability to obtain funds necessary for
the activities for which the permit or
license is sought.
§ 53.1680

Annual Financial Reports

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This proposed rule would add
§ 53.1680, which would require
licensees and holders of a CP under part
53 to submit annual financial reports to
the Commission, with exceptions for
those that submit financial forms to the
Securities and Exchange Commission or
the Federal Energy Regulatory
Commission.
§ 53.1690 Licensee’s Change of Status;
Financial Qualifications
This proposed rule would add
§ 53.1690, which would require electric
utility licensees that hold an OL or COL
for a commercial nuclear plant under
part 53 to provide the NRC with the

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Creditor Regulations

This proposed rule would add
§ 53.1700, which would establish
regulations with respect to the creditors
of any facility under part 53.
§ 53.1710

Financial Protection

This proposed rule would add
§ 53.1710, which would establish
requirements for licenses under part 53
to obtain and maintain insurance to
cover the costs of an accident.
§ 53.1720 Insurance Required To
Stabilize and Decontaminate Plant
Following an Accident

§ 70.24 Criticality Accident
Requirements

§ 70.32

Conditions of Licenses

This proposed rule would revise
§ 70.32(c)(1) and (d) to incorporate the
appropriate references to part 53.
§ 70.50

Reporting Requirements

This proposed rule would revise
§ 70.50(d) to clarify the applicability of
the reporting requirements of this
section to part 53 licensees.
§ 72.3

Definitions

This proposed rule would add
§ 53.1720, which would require
commercial nuclear plant licensees
under part 53 to obtain insurance
sufficient to cover the costs of
stabilizing and decontaminating the
plant in the event of an accident.

This proposed rule would revise the
definition of ‘‘Independent spent fuel
storage installation or ISFSI’’ in § 72.3 to
include a reference to facilities licensed
under part 53.

§ 53.1730 Financial Protection
Requirements

This proposed rule would revise
§ 72.30(e)(5) to include the appropriate
references to part 53.

This proposed rule would add
§ 53.1730, which would require
commercial nuclear plant licensees
under part 53 to satisfy the provisions
of part 140.
Subpart M—Enforcement
This proposed rule would add subpart
M, which would address certain
violations and penalties associated with
violations of part 53 regulations.
§ 53.9000

Violations

This proposed rule would add
§ 53.9000, providing notice of the
Commission’s authority to obtain
injunctions or other court orders for the
violations enumerated in this section.
§ 53.9010

Criminal Penalties

This proposed rule would add
§ 53.9010, providing notice to all
persons and entities subject to part 53
that they are subject to criminal
sanctions for willful violations,
attempted violations, or conspiracy to
violate certain regulations under part
53.
§ 70.20a General License to Possess
Special Nuclear Material for Transport
This proposed rule would revise
§ 70.20a(b) to include a reference to part
53.
§ 70.22

Contents of Applications

This proposed rule would revise
§ 70.22, paragraphs (b), (h)(1), (j)(1), and

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§ 72.30 Financial Assurance and
Recordkeeping for Decommissioning

§ 72.32

Emergency Plan

This proposed rule would revise
§ 72.32(c)(2) to include a reference to
the exclusion area as defined in part 53.
§ 72.40

Issuance of License

This proposed rule would revise
§ 72.40(c) regarding the issuance of a
license under part 72 to include a
reference to previous licensing actions,
including the issuance of a CP under
part 53.
§ 72.75 Reporting Requirements for
Specific Events and Conditions
This proposed rule would revise
§ 72.75(i)(1)(ii) regarding reporting
requirements for specific events and
conditions with references to reactors
licensed under part 53.
§ 72.184

Safeguards Contingency Plan

This proposed rule would revise
§ 72.184(a) regarding the requirements
of a licensee’s safeguarding contingency
plan with a reference to nuclear
facilities licensed under part 53.
§ 72.210

General License Issued

This proposed rule would revise
§ 72.210 to issue a general license for
the storage of spent fuel in an
independent spent storage installation
at power to persons authorized to
possess or operate nuclear power
reactors under part 53.

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Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules
§ 72.212 Conditions of General License
Issued Under § 72.210
This proposed rule would revise
§ 72.212(b)(8) regarding the conditions
of a general license issued under
§ 72.210 to include a reference to license
amendments for a facility made
pursuant to part 53.
§ 72.218 Termination of Licenses
This proposed rule would revise
§ 72.218(a) to include a reference to the
notification required under part 53
regarding the plan for managing spent
fuel prior to decommissioning. It would
also extend the provisions of § 72.218(b)
to a reactor operating or COL under part
53.
§ 73.1 Purpose and Scope
This proposed rule would revise
§ 73.1(b)(1)(i) to extend the scope of part
73 to production and utilization
facilities licensed under part 53, in
addition to parts 50 and 52.

§ 73.57 Requirements for Criminal
History Records Checks of Individuals
Granted Unescorted Access to a Nuclear
Power Facility, a Non-power Reactor, or
Access to Safeguards Information
This proposed rule would revise
§ 73.57(a)(3) to incorporate the
appropriate references to OLs granted
under part 53 and Commission findings
under § 53.1452(g) regarding the
requirement for license applicants to
submit fingerprints for all personnel
with unescorted access.
§ 73.58 Safety/Security Interface
Requirements for Nuclear Power
Reactors
This proposed rule would revise
§ 73.58(a) to extend the requirements of
this section to part 53 licensees.

§ 73.2 Definitions
This proposed rule would revise
§ 73.2 introductory text and paragraph
(a) such that terms defined in part 53
have the same meaning in part 73.

§ 73.67 Licensee Fixed Site and InTransit Requirements for the Physical
Protection of Special Nuclear Material
of Moderate and Low Strategic
Significance

§ 73.8 Information Collection
Requirements: OMB Approval
This proposed rule would revise
§ 73.8(b) with the new information
collection requirements contained in
proposed §§ 73.77, 73.100, 73.110, and
73.120.

This proposed rule would revise
§ 73.67(d) and (f) to include a reference
to licensees authorized to operate a
nuclear power plant under part 53.

§ 73.50 Requirements for Physical
Protection of Licensed Activities
This proposed rule would revise
§ 73.50 to exempt nuclear reactor
facilities licensed under part 53, in
addition to parts 50 and 52, from the
requirements of this section.

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under part 53 who do not demonstrate
compliance with certain requirements
under part 53.

§ 73.56 Personnel Access
Authorization Requirements for Nuclear
Power Plants
This proposed rule would revise
§ 73.56(a)(3) to apply this section’s
personnel AA requirements to
applicants for an OL or holders of a COL

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This proposed rule would revise
§ 73.77, paragraphs (a), (b), (c)(6) and (7)
regarding the notification process for
cybersecurity events to include
notifications for the declaration of an
emergency class made under part 53.
Subpart J—Security Requirements at
Commercial Nuclear Plants

§ 73.55 Requirements for Physical
Protection of Licensed Activities in
Nuclear Power Reactors Against
Radiological Sabotage
This proposed rule would revise
§ 73.55, paragraphs (a)(4) and (6),
(i)(4)(iii), (l)(1), (l)(7)(ii), (p)(1)(i), (r)(2),
and (r)(4)(iii), to incorporate the
appropriate references to part 53
regarding requirements for physical
protection of licensed activities in
nuclear power reactors against
radiological sabotage.

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§ 73.77 Cybersecurity Event
Notifications

This proposed rule would add new
Subpart J of part 73 containing
§§ 73.100, 73.110, and 73.120, to
establish security requirements for
commercial nuclear plants licensed
under part 53.
§ 73.100 Technology-Inclusive
Requirements for Physical Protection of
Licensed Activities at Commercial
Nuclear Plants Against Radiological
Sabotage
This proposed rule would add
§ 73.100, which would establish a
performance-based regulatory
framework for physical protection as an
alternative to the prescriptive
requirements of § 73.55, which also
governs physical protection programs
for part 50 and 52 licensees.

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§ 73.110 Technology-Inclusive
Requirements for Protection of Digital
Computer and Communication Systems
and Networks
This proposed rule would add
§ 73.110, which would establish a
consequence-based approach to
cybersecurity and would require that
part 53 licensees demonstrate
reasonable assurance that digital
computer and communication systems
and networks are adequately protected
against cyberattacks in a manner that is
commensurate with the potential
consequences of those attacks.
§ 73.120 Access Authorization
Program for Commercial Nuclear Plants
This proposed rule would add
§ 73.120, which would establish
performance objectives as an alternative
to compliance with the AA provisions
of §§ 73.55, 73.56, and 73.57. This
proposed rule would afford part 53
licensees additional flexibility in
establishing an AA program that
demonstrates compliance with the
performance objectives and
requirements of this section.
§ 73.1200 Notification of Physical
Security Events
This proposed rule would revise
§ 73.1200, paragraphs (a), (c)(1), (e)(1),
(e)(3), (e)(4), (g)(1), (o)(5)(i), (o)(6)(i), (r),
and (s) to extend the requirements of
this section to part 53 licensees.
§ 73.1205 Written Follow-Up Reports
of Physical Security Events
This proposed rule would revise
§ 73.1205(b)(2) to extend the
requirements of this section to part 53
licensees.
§ 73.1210 Recordkeeping of Physical
Security Events
This proposed rule would revise
§ 73.1210(a)(1) and (b)(3)(i) to extend
the requirements of this section to part
53 licensees.
§ 73.1215 Suspicious Activity Reports
This proposed rule would revise
§ 73.1215(d)(1) to include a reference to
§ 73.100.
Appendix B to part 73—General Criteria
for Security Personnel
This proposed rule would revise
appendix B to part 73 to state that terms
defined in part 53 have the same
meaning when used in this appendix.
§ 74.31 Nuclear Material Control and
Accounting for Special Nuclear Material
of Low Strategic Significance
This proposed rule would revise
§ 74.31(a) to include a reference to

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Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules
§ 140.12 Amount of Financial
Protection Required for Other Reactors

production or utilization facilities
licensed under part 53, in addition to
parts 50 and 70.
§ 74.41 Nuclear Material Control and
Accounting for Special Nuclear Material
of Moderate Strategic Significance
This proposed rule would revise
§ 74.41(a) to include a reference to
nuclear reactors licensed under part 53.
§ 74.51 Nuclear Material Control and
Accounting for Strategic Special
Nuclear Material
This proposed rule would revise
§ 74.51(a) to include a reference to
nuclear reactors licensed under part 53.
§ 75.4

Definitions

This proposed rule would revise
§ 75.4 such that terms defined in
§ 53.020 have the same meaning when
used in this part. The definition of
‘‘Facility’’ would also be revised to
include any plant or location where
more than 1 effective kilogram of
nuclear material is licensed pursuant to
part 53.
§ 95.5

Definitions

This proposed rule would revise the
definition of ‘‘License’’ in § 95.5 to
include those issued under part 53.
§ 95.39 External Transmission of
Documents and Material
This proposed rule would revise
§ 95.39(a) to apply restrictions to the
external transmission of documents and
material containing classified
information in connection with NRC
licenses, certificates, standard design
approvals, or standard DCs issued under
part 53.
§ 140.2

Scope

This proposed rule would revise
§ 140.2(a)(1) and (2) to include part 53
applicants and licensees within the
scope of part 140 regulations.
§ 140.10

Scope

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This proposed rule would revise
§ 140.10 to apply the provisions of
subpart B to applicants or holders of a
license to operate a nuclear reactor
under part 53, as well as applicants and
holders of a COL under part 53.

This proposed rule would revise
§ 140.12(c) to require the licensee’s
primary financial protection to cover all
reactors in any case where a person is
authorized under part 53 to operate two
or more nuclear reactors at the same
location.
§ 140.13 Amount of Financial
Protection Required of Certain Holders
of Construction Permits and Combined
Licenses Under 10 CFR Part 52
This proposed rule would revise
§ 140.13 with the appropriate references
to part 53 regarding the requirement for
holders of a CP or COL under part 53
to obtain financial protection.
§ 140.20
Liens

Indemnity Agreements and

This proposed rule would revise
§ 140.20(a)(1)(i) and (ii) with
appropriate references to part 53.
§ 150.15

Persons Not Exempt

The proposed rule would revise
§ 150.15, paragraphs (a)(7)(iii) and (a)(8)
to add a reference to facilities licensed
under parts 53 and 52.
§ 170.3

Definitions

The proposed rule would revise
§ 170.3 to incorporate references to part
53 into the definitions of
‘‘Manufacturing license,’’ ‘‘Part 55
Reviews,’’ ‘‘Power reactor,’’ and
‘‘Special projects.’’
§ 170.12

Payment of Fees

The proposed rule would revise
§ 170.12(d)(1)(v) regarding special
project fees in connection with FSARs
to include part 53.
§ 170.21 Schedule of Fees for
Production and Utilization Facilities,
Review of Standard Referenced Design
Approvals, Special Projects,
Inspections, And import and Export
Licenses
The proposed rule would revise
§ 170.21, footnote 1 to include fees
charged for approvals issued under the
exemption provision in § 53.080.

§ 140.11 Amounts of Financial
Protection for Certain Reactors

§ 170.41 Failure by Applicant or
Licensee to Pay Prescribed Fees

This proposed rule would revise
§ 140.11(b) to require the licensee’s
primary financial protection to cover all
reactors in any case where a person is
authorized under part 53 to operate two
or more nuclear reactors at the same
location.

The proposed rule would revise
§ 170.41 to include a general reference
to part 53 in connection with remedial
actions that the Commission might take
when an applicant or licensee fails to
pay a prescribed fee required by this
part.

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§ 171.3

Scope

The proposed rule would revise
§ 171.3 to apply the provisions of this
part to any person holding an OL for a
power reactor licensed under part 53 or
a COL issued under part 53.
§ 171.5

Definitions

This proposed rule would revise the
definitions of ‘‘Operating license’’ and
‘‘Power reactor’’ in § 171.5 to
incorporate the appropriate references
to part 53.
§ 171.15 Annual fees: Non-Power
Production or Utilization Licenses,
Reactor Licenses, and Independent
Spent Fuel Storage Licenses
This proposed rule would revise
§ 171.15, paragraphs (a), (b)(2)(iii),
(c)(1), and (d)(1) regarding annual fees
that are applicable to part 53 licensees.
§ 171.17

Proration

This proposed rule would revise
§ 171.17, paragraphs (a), (a)(1)(ii) and
(a)(2) with references to part 53 licenses.
VIII. Regulatory Flexibility
Certification
The Regulatory Flexibility Act of
1980, as amended at 5 U.S.C. 601 et seq,
requires that agencies consider the
impact of their rulemakings on small
entities and, consistent with applicable
statutes, consider alternatives to
minimize these impacts on the
businesses, organizations, and
government jurisdictions to which they
apply.
In accordance with the Small
Business Administration’s (SBA’s)
regulation at 13 CFR 121.903(c), the
NRC has developed its own size
standards for performing an RFA
analysis and has verified with the SBA
Office of Advocacy that its size
standards are appropriate for NRC
analyses. The NRC size standards at
§ 2.810, ‘‘NRC size standards,’’ are used
to determine whether an applicant or
licensee qualifies as a small entity in the
NRC’s regulatory programs. Section
2.810 defines the following types of
small entities:
Small business is a for-profit concern
and is a—(1) Concern that provides a
service or a concern not engaged in
manufacturing with average gross
receipts of $8.0 million or less over its
last 5 completed fiscal years; or (2)
Manufacturing concern with an average
number of 500 or fewer employees
based upon employment during each
pay period for the preceding 12 calendar
months.
Small organization is a not-for-profit
organization which is independently

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Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules
owned and operated and has annual
gross receipts of $8.0 million or less.
Small governmental jurisdiction is a
government of a city, county, town,
township, village, school district, or
special district with a population of less
than 50,000.
Small educational institution is one
that is—(1) Supported by a qualifying
small governmental jurisdiction; or (2)
Not State or publicly supported and has
500 or fewer employees.
Number of Small Entities Affected
The NRC is currently not aware of any
known small entities as defined in
§ 2.810 that are planning to apply for a
commercial nuclear plant ESP, CP, OL,
ML, or COL under part 53 that would be
impacted by this proposed rule. Based
on this finding, the NRC has
preliminarily determined that the
proposed rule would not have a
significant economic impact on a
substantial number of small entities.

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Economic Impact on Small Entities
Depending on how the ownership
and/or operating responsibilities for
such an enterprise were structured,
applicants for a commercial nuclear
plant rated 8 Megawatts electric (MWe)
or less could conceivably qualify as
small entities as defined by § 2.810.
Owners that operate power reactors
rated greater than 8 MWe could generate
sufficient electricity revenue that
exceeds the gross annual receipts limit
of $8 million, assuming a 90 percent
capacity factor and the June 2021 DOE’s
Energy Information Administration U.S.
average price of electricity to the
ultimate customer for all sectors of 11.3
cents per kilowatt-hour.
Although the NRC is not aware of any
small entities that would be affected by
the proposed rule, there is a possibility
that future applications for a
commercial nuclear plant permit or
license could be submitted by small
entities who plan to own and operate a
commercial nuclear plant rated 8 MWe
or less. Commercial nuclear plants that
are rated 8 MWe or less would most
likely be used to support electrical
demand for military bases or small
remote towns and would provide
process heat, so they would not directly
compete with a larger commercial
nuclear plant that would typically
produce electricity for the grid. As a
result of these differing purposes, the
NRC would expect that small and large
entities would not be in direct
competition with each other.
Therefore, the NRC preliminarily
concludes that this proposed rule would
not have a significant economic impact

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on a substantial number of small
entities.
Request for Comments
The NRC is seeking comment on both
its initial RFA analysis and on its
preliminary conclusion that this
proposed rule would not have a
significant economic impact on a
substantial number of small entities
because of the likelihood that most
expected applicants would not qualify
as a small entity. Additionally, the NRC
is seeking comment on its preliminary
conclusion that if a small entity were to
submit a commercial nuclear plant
application, the small entity would not
incur a significant economic impact as
it would most likely not be in
competition with a large entity.
Any small entity that could be subject
to this regulation that determines,
because of its size, it is likely to bear a
disproportionate adverse economic
impact should notify the Commission of
this opinion in a comment that
indicates—
1. The applicant’s size and how the
proposed regulation would impose a
significant economic burden on the
applicant as compared to the economic
burden on a larger applicant;
2. How the proposed regulations
could be modified to take into account
the applicant’s differing needs or
capabilities;
3. The benefits that would accrue or
the detriments that would be avoided if
the proposed regulations were modified
as suggested by the applicant;
4. How the proposed regulation, as
modified, would more closely equalize
the impact of NRC regulations or create
more equal access to the benefits of
Federal programs as opposed to
providing special advantages to any
individual or group; and
5. How the proposed regulation, as
modified, would still adequately
demonstrate compliance with the NRC’s
obligations under the Act.
IX. Regulatory Analysis
The NRC has prepared a draft
regulatory analysis for this proposed
rule. The analysis examines the costs
and benefits of the alternatives
considered by the NRC. The conclusion
from the analysis is that this proposed
rule and associated guidance would
result in net averted costs to the
industry and the NRC of $28.1 million
using a 7-percent discount rate and
$34.5 million using a 3-percent discount
rate due to reductions in exemption
requests. The analysis also assumes one
applicant under part 53. As the number
of applicants increases, so do the
estimated averted costs. The NRC

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requests public comment on the draft
regulatory analysis, which is available
as indicated in the ‘‘Availability of
Documents’’ section of this document.
Comments on the draft regulatory
analysis may be submitted to the NRC
as indicated under the ADDRESSES
caption of this document.
X. Backfitting and Issue Finality
This section describes the backfitting
and issue finality implications of this
proposed rule and the draft guidance
documents described in section XVIII,
‘‘Availability of Guidance,’’ in this
document, as applied to pertinent NRC
approvals and certain applicants that
reference NRC approvals in their
applications. The NRC’s current
backfitting provisions associated with
nuclear power plants appear in § 50.109,
‘‘Backfitting,’’ and apply to CPs and OLs
under part 50. Issue finality provisions
(analogous to the backfitting provisions
in § 50.109) for approvals under part 52
are located in various provisions of part
52. The NRC Management Directive 8.4,
‘‘Management of Backfitting, Forward
Fitting, Issue Finality, and Information
Requests,’’ describes the Commission’s
policies on backfitting and issue
finality.
This proposed rule would provide a
regulatory scheme for entities to apply
for approvals under part 53. The part 50
backfitting provisions and part 52 issue
finality provisions apply to actions
taken by the NRC under part 50 or part
52, respectively, or actions taken by the
NRC under other parts of 10 CFR
chapter I that, for holders of certain
approvals under part 50 or part 52,
inextricably affect their activities
regulated under part 50 or part 52.
Issuance and implementation of
proposed part 53 would not constitute
actions taken under part 50 or part 52.
Also, proposed part 53 would not allow
an applicant to reference approvals
issued under part 50 or part 52.
Therefore, the issuance and
implementation of proposed part 53
would not affect part 50 or part 52
entities’ activities regulated under part
50 or part 52. Therefore, the addition of
part 53 through this proposed rule
would not be within the scope of the
part 50 backfitting and part 52 issue
finality provisions.
The NRC also proposes conforming
changes to parts 1, 2, 10, 11, 19, 20, 21,
25, 26, 30, 40, 50, 51, 70, 72, 73, 74, 75,
95, 140, 150, 170, and 171 to reflect the
addition of part 53. These changes
would not meet the definition of
‘‘backfitting’’ in § 50.109 or § 70.76,
‘‘Backfitting,’’ because the proposed
changes would not modify or add to the
systems, structures, components, or

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design of a facility or to the procedures
or organization required to operate a
facility under part 50 or 70. These
changes would not meet the definition
of ‘‘backfitting’’ in § 72.62,
‘‘Backfitting,’’ because the proposed
changes would not add, eliminate, or
modify the SSCs of an independent
spent fuel storage installation (ISFSI) or
the procedures or organization required
to operate an ISFSI. These proposed
changes would not inextricably affect
activities regulated under parts 50, 52,
70, or 72. Therefore, the proposed
changes to parts 1, 2, 10, 11, 19, 20, 21,
25, 26, 30, 40, 50, 51, 70, 72, 73, 74, 75,
95, 140, 150, 170, and 171 would not
constitute backfitting under parts 50, 70,
or 72 or affect the issue finality of an
approval under part 52.
The NRC is issuing 10 draft guidance
documents that, if issued as final
guidance documents, would provide
guidance on the methods acceptable to
the NRC for complying with aspects of
this proposed rule. These documents
would not apply to holders of approvals
issued under part 50 or part 52. Further,
as discussed in the guidance
documents, applicants and licensees
would not be required to comply with
the positions set forth in the guidance.
Therefore, issuance of the guidance
documents as final guidance would not
constitute backfitting under part 50 or
affect the issue finality of any approval
issued under part 52.

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XI. Cumulative Effects of Regulation
The NRC seeks to minimize any
potential negative consequences
resulting from the cumulative effects of
regulation (CER). The CER describes the
challenges that licensees, or other
impacted entities such as State partners,
may face while implementing new
regulatory positions, programs, or
requirements (e.g., rules, generic letters,
backfits, inspections). The CER is an
organizational effectiveness challenge
that may result from a licensee or
impacted entity implementing a number
of complex regulatory actions,
programs, or requirements within
limited available resources. The NRC’s
CER process involved engaging with
external stakeholders throughout this
proposed rule and related regulatory
activities. Public involvement has
included numerous public meetings to
examine the part 53 risk-informed,
technology-inclusive requirements for
commercial nuclear plants and the
publication of numerous versions of
preliminary proposed rule language.
The NRC is considering holding
additional public meetings during the
remainder of the rulemaking process.

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In parallel with this proposed rule,
the NRC is issuing 10 draft
implementing guidance documents for
comment to support informed external
stakeholder feedback. Section XVII,
‘‘Availability of Guidance,’’ of this
document describes how the public can
access the draft implementing guidance.
In addition to the questions in the
‘‘Specific Requests for Comments’’
section of this document, the NRC is
requesting CER feedback on the
following questions:
1. In light of any current or projected
CER challenges, does the proposed
rule’s effective date provide sufficient
time to implement the new proposed
requirements, including changes to
programs, procedures, and the facility?
2. If CER challenges currently exist or
are expected, what should be done to
address them? For example, if more
time is required for implementation of
the new requirements, what period of
time is sufficient?
3. Do other (NRC or other agency)
regulatory actions (e.g., orders, generic
communications, license amendment
requests, inspection findings of a
generic nature) influence the
implementation of the proposed rule’s
requirements?
4. Are there unintended
consequences? Does the proposed rule
create conditions that would be contrary
to the proposed rule’s purpose and
objectives? If so, what are the
unintended consequences, and how
should they be addressed?
5. Please comment on the NRC’s cost
and benefit estimates in the regulatory
analysis that supports this proposed
rule. The draft regulatory analysis is
available as indicated under the
‘‘Availability of Documents’’ section of
this document.
XII. Plain Writing
The Plain Writing Act of 2010 (Pub.
L. 111–274) requires Federal agencies to
write documents in a clear, concise, and
well-organized manner. The NRC has
written this document to be consistent
with the Plain Writing Act as well as the
Presidential Memorandum, ‘‘Plain
Language in Government Writing,’’
published June 10, 1998 (63 FR 31885).
The NRC requests comment on this
document with respect to the clarity and
effectiveness of the language used.
XIII. Environmental Assessment and
Proposed Finding of No Significant
Environmental Impact
The Commission has preliminarily
determined under the National
Environmental Policy Act of 1969, as
amended, and the Commission’s
regulations in subpart A of part 51, that

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this rule, if adopted, would not be a
major Federal action significantly
affecting the quality of the human
environment, and an EIS is not required.
The implementation of the proposed
rule requirements does not have a
significant impact on the environment.
The proposed rulemaking would either
have requirements that are
administrative in application, matters of
procedure, or provide an equivalent
level of safety as existing requirements;
therefore, there would be similar
environmental impacts from the
implementation of the part 53
regulations as there are for existing
requirements.
The preliminary determination of this
EA is that there will be no significant
effect on the quality of the human
environment from this action. Public
stakeholders should note, however, that
comments on any aspect of this EA may
be submitted to the NRC as indicated
under the ADDRESSES section of this
document. The EA is available as
indicated under the ‘‘Availability of
Documents’’ section of this document.
The NRC has sent a copy of the EA,
and this proposed rule to every State
Liaison Officer and has requested
comments.
XIV. Paperwork Reduction Act
This proposed rule contains new
collections of information contained in
parts 26, 50, 53, and 73 and NRC Forms
361S, 366, 366A, 366B, 893, and 894
that are subject to the Paperwork
Reduction Act of 1995 (44 U.S.C. 3501
et seq). The collections of information
have been submitted to the OMB for
review and approval. The proposed
changes to parts 2, 10, 11, 19, 20, 21, 25,
30, 40, 51, 70, 72, 74, 75, 95, 140, 150,
170, and 171 do not contain any new or
amended collections of information
subject to the Paperwork Reduction Act
of 1995. Existing collections of
information were approved by the OMB,
approval numbers 3150–0062 (part 11),
3150–0044 (part 19), 3150–0014 (part
20), 3150–0035 (part 21), 3150–0046
(part 25), 3150–0017 (part 30), 3150–
0020 (part 40), 3150–0021 (part 51),
3150–0024 (NRC Form 396), 3150–0090
(NRC Form 398), 3150–0009 (part 70),
3150–0132 (part 72), 3150–0123 (part
74), 3150–0055 (part 75), 3150–0047
(part 95), 3150–0039 (part 140), and
3150–0032 (part 150).
Type of submission, new or revision:
Revision and new.
The title of the information collection:
Risk-Informed, Technology-Inclusive
Regulatory Framework for Advanced
Reactors.

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The form number if applicable: NRC
Forms 361S, 366, 366A, 366B, 893, and
894.
How often the collection is required or
requested: Once, on occasion, every 30
days, biannually, annually, biennially,
every four years, every five years, every
ten years.
Who will be required or asked to
respond: Part 53 commercial nuclear
plant licensees and license applicants
for commercial nuclear plants to be
licensed under part 53.
An estimate of the number of annual
responses: 15 (2 responses for Part 26,
11 responses for Part 53, 2 responses for
Part 50 and 0 responses for Part 73 and
NRC Forms 361S, 366, 366A, 366B, 893,
and 894)
The estimated number of annual
respondents: 2 (2 respondents for Part
26, 2 respondents for Part 53, 2
respondents for Part 50 and 0
respondents for Part 73 and NRC Forms
361S, 366, 366A, 366B, 893, and 894)
An estimate of the total number of
hours needed annually to comply with
the information collection requirement
or request: 230,244 hours. (656 hours for
Part 26, 220,801 hours for Part 53, 8,767
hours for Part 50 and 0 hours for Part
73 and NRC Forms 361S, 366, 366A,
366B, 893, and 894)
Abstract: The NRC is proposing to
establish an optional technologyinclusive regulatory framework for use
by applicants for new commercial
nuclear plant designs. The regulatory
requirements developed in this
rulemaking would use methods of
evaluation, including risk-informed and
performance-based methods, that are
flexible and practicable for application
to a variety of new reactor technologies.
The NRC’s goals in amending these
regulations are to continue to provide
reasonable assurance of adequate
protection of public health and safety
and the common defense and security at
reactor sites at which new nuclear
reactor designs are deployed to at least
the same degree of protection as
required for current-generation LWRs;
protect health and minimize danger to
life or property to at least the same
degree of protection as required for
current-generation LWRs; provide
greater operational flexibilities where
supported by enhanced margins of
safety that may be provided in new
nuclear designs; and promote regulatory
stability, predictability, and clarity.
The proposed rule covers diverse
topics, which result in recordkeeping
and reporting requirements related to
contents of applications, plant design
and analysis, siting, construction and
manufacturing, licensing-basis
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programs, staffing, FFD, physical
security, cyber-security, AA,
decommissioning, and quality
assurance.
In addition to the new information
collections in the proposed regulations,
part 53 would result in new collections
via NRC Forms 361S, 366, 366A, 366B,
893, and 894. NRC Forms 366, 366A,
and 366B would be modified to include
part 53 reportable events covering an
equivalent scope as the requirements in
10 CFR 50.73, but without LWR-specific
terminology to ensure technology
inclusiveness. The proposed rule also
would require part 53 licensees to use
NRC Forms 893 and 894 to report on
positive drug and alcohol test results
(NRC Form 893) and annual fitness-forduty program performance (NRC Form
894). Finally, a new version of NRC
Form 361 (NRC Form 361S) would be
created for use by part 53 licensees,
covering an equivalent scope as the
requirements in 10 CFR 50.72, but
without LWR-specific terminology to
ensure technology inclusiveness.
The NRC is seeking public comment
on the potential impact of the
information collections contained in
this proposed rule and on the following
issues:
1. Is the proposed information
collection necessary for the proper
performance of the functions of the
NRC, including whether the information
will have practical utility? Please
explain your response.
2. Is the estimate of the burden of the
proposed information collection
accurate? Please explain your response.
3. Is there a way to enhance the
quality, utility, and clarity of the
information to be collected? Please
explain your response.
4. How can the burden of the
proposed information collection on
respondents be minimized, including
the use of automated collection
techniques or other forms of information
technology? Please explain your
response.
The OMB clearance documents and
proposed rule is available as indicated
under the ‘‘Availability of Documents’’
section in this document or may be
viewed free of charge by contacting the
NRC’s PDR reference staff at 1–800–
397–4209, at 301–415–4737, or by email
to PDR.resource@nrc.gov. You may
obtain information and comment
submissions related to the OMB
clearance package by searching on
http://www.regulations.gov under
Docket ID NRC–2019–0062.
You may submit comments on any
aspect of these proposed information
collections, including suggestions for

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reducing the burden and on the above
issues, by the following methods:
• Federal Rulemaking Website: Go to
http://www.regulations.gov and search
for Docket ID NRC–2019–0062.
• Mail comments to: FOIA, Library,
and Information Collections Branch,
Office of the Chief Information Officer,
Mail Stop: T6–A10M, U.S. Nuclear
Regulatory Commission, Washington,
DC 20555–0001 or by email to
Infocollects.Resource@nrc.gov or to the
OMB reviewer at: OMB Office of
Information and Regulatory Affairs
(3150–XXXX, 3150–0002, –0104, –0146,
–0238), Attn: Desk Officer for the
Nuclear Regulatory Commission, 725
17th Street NW, Washington, DC 20503.
Submit comments by December 2,
2024. Comments received after this date
will be considered if it is practical to do
so, but the NRC staff is able to ensure
consideration only for comments
received on or before this date.
Public Protection Notification
The NRC may not conduct or sponsor,
and a person is not required to respond
to, a collection of information unless the
document requesting or requiring the
collection displays a currently valid
OMB control number.
XV. Criminal Penalties
For the purposes of Section 223 of the
Act, the NRC is issuing this proposed
rule that would add a new part 53 and
amend parts 26 and 73 under one or
more of Sections 161b, 161i, or 161o of
the Act, except as noted in proposed
§ 53.9010(b) and § 26.825(b). Willful
violations of the part 53 and part 26
regulations not listed in proposed
§ 53.9010(b) and § 26.825(b) would be
subject to criminal enforcement.
Criminal penalties as they apply to
regulations in part 53 would be
discussed in § 53.9010.
XVI. Voluntary Consensus Standards
The NTTAA requires that Federal
agencies use technical standards that are
developed or adopted by voluntary
consensus standards bodies unless the
use of such a standard is inconsistent
with applicable law or otherwise
impractical. In this proposed rule, the
NRC would revise regulations by adding
a risk-informed, technology-inclusive
regulatory framework for commercial
advanced nuclear reactors. This action
does not constitute the establishment of
a standard that contains generally
applicable requirements.
XVII. Availability of Guidance
As discussed in section II,
Background, of this document, the
NRC’s development of proposed part 53

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built upon recent and ongoing activities
such as those described in SECY–19–
0117. Because a number of those
activities are ongoing to support new
reactor applications under the existing
regulatory framework of 10 CFR parts 50
and 52, the NRC staff identified in its
response to SRM–SECY–20–0032 that
the timing of guidance document
development to support the part 53
rulemaking was a key risk and
uncertainty to publishing the final part
53 rule. To mitigate this risk, the NRC
engaged external stakeholders to ensure
a common prioritization of the
development of these guidance
documents and to work diligently on
those that would be needed to support
this rulemaking, forthcoming
applications, or broader efforts such as
the Advanced Reactor Demonstration
Program being sponsored by the DOE.
The NRC also recognizes that guidance
development to support part 53 and
advanced reactors will continue as the
industry and NRC learn lessons from
licensing reviews and operating
experience. Therefore, the NRC
categorized guidance supporting the
part 53 rulemaking into three categories:
(1) guidance issued or under
development to support applications
under the existing regulatory
framework; (2) implementing guidance
for part 53-specific proposed rule
language; and (3) future guidance
activities that would need to be
completed after the part 53 proposed
rule is published for public comment.
(1) Hundreds of guidance documents
exist for the current fleet of operating
reactors. While some of the guidance is
specific to LWR technologies, other
guidance is technology inclusive in
nature and should be considered, as
appropriate, in the development of all
licensing applications and NRC reviews.
In addition, the NRC has undertaken
efforts to incorporate or reference the
most relevant guidance in its efforts to
develop additional guidance for future
advanced reactors. The NRC has issued
the following guidance to support
licensing reviews of advanced reactors
under the existing regulatory framework
that will continue to inform applicant
development and NRC reviews under
parts 50 and 52. Conforming changes to
these guidance documents would be
needed to ensure they are applicable
under part 53. The NRC will issue
revisions or part 53-related companions
to these guidance documents for public
comment after the publication of this
proposed rule and then finalize and
issue the guidance documents with or
after the final part 53 rule.

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• RG 1.233, ‘‘Guidance for a
Technology-Inclusive, Risk-Informed,
and Performance-Based Methodology
to Inform the Licensing Basis and
Content of Applications for Licenses,
Certifications, and Approvals for NonLight-Water Reactors’’
• RG 1.247 for trial use, ‘‘Acceptability
of Probabilistic Risk Assessment
Results for Non-Light-Water Reactor
Risk-Informed Activities’’
• NUREG–2246, ‘‘Fuel Qualification for
Advanced Reactors’’
• RG 1.87, Revision 2, ‘‘Acceptability of
ASME Code, Section III, Division 5,
‘‘High Temperature Reactors’’
• RG 1.246, ‘‘Acceptability of ASME
Code, Section XI, Division 2,
‘Requirements for Reliability And
Integrity Management (RIM) Programs
for Nuclear Power Plants,’ for NonLight Water Reactors’’
Also, the NRC continues to develop
additional guidance to support licensing
reviews of advanced reactors under the
existing regulatory framework. Some of
these guidance documents have been
issued and others will be issued before
the finalization of part 53 to support
near-term applicants and NRC reviews.
For example, the NRC has been and
continues to be engaged with the DOE
and industry to develop content of
application guidance and other
regulatory guidance for advanced
reactors to support applications and
subsequent operations under the
existing regulatory framework. These
guidance documents, such as the
industry-led Technology-Inclusive
Content of Application Project guidance
found in NEI 21–07, Revision 1, and the
NRC-led Advanced Reactor Content of
Application Project (ARCAP) interim
staff guidance (ISG) documents and
NRC regulatory guidance endorsing NEI
21–07, Revision 1, will support
developers in preparing advanced
reactor applications. These guidance
documents provide an overview of the
information that should be included in
an advanced reactor application, a
review roadmap for the NRC with the
principal purpose of ensuring
consistency, quality, and uniformity of
NRC reviews, and a well-defined base
from which the NRC can evaluate
proposed changes in the scope and
requirements of reviews. While specific
sections of the information are primarily
aligned with the LMP methodology, as
endorsed in RG 1.233, as one acceptable
process for applicants to use when
developing portions of an application,
the concepts and general information
may be used to inform the review of an
application submitted using other
traditional licensing approach

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methodologies (as applicable). Other
sections of the information are generally
applicable and independent of the
methodology used to develop an
advanced reactor application. The
ARCAP ISGs provide references to
numerous regulatory guidance
documents that should be considered by
both applicants and the NRC in
developing and reviewing, respectively,
advanced reactor applications. The NRC
has issued the following documents
separately from this proposed rule. The
NRC may issue other, related guidance
documents with or after the final part 53
rule.
• RG 1.253, ‘‘Guidance for a Technology
Inclusive Content of Application
Methodology to Inform the Licensing
Basis and Content of Applications for
Licenses, Certifications, and
Approvals for Non-Light-Water
Reactors’’
• DANU–ISG–2022–01, ‘‘Advanced
Reactor Content of Application
Project, ‘Review of Risk-Informed,
Technology-Inclusive Advanced
Reactor Applications—Roadmap’ ’’
• DANU–ISG–2022–02, ‘‘Advanced
Reactor Content of Application
Project Chapter 2, ‘Site Information’ ’’
• DANU–ISG–2022–03, ‘‘Advanced
Reactor Content of Application
Project Chapter 9, ‘Control of Routine
Plant Radioactive Effluents, Plant
Contamination and Solid Waste’ ’’
• DANU–ISG–2022–04, ‘‘Advanced
Reactor Content of Application
Project Chapter 10, ‘Control of
Occupational Dose’ ’’
• DANU–ISG–2022–05, ‘‘Advanced
Reactor Content of Application
Project Chapter 11, ‘Organization and
Human-System Considerations’ ’’
• DANU–ISG–2022–06, ‘‘Advanced
Reactor Content of Application
Project Chapter 12, ‘Post-Construction
Inspection, Testing, and Analysis
Program’ ’’
• DANU–ISG–2022–07, ‘‘Advanced
Reactor Content of Application
Project, ‘Risk-Informed Inservice
Inspection/Inservice Testing’ ’’
• DANU–ISG–2022–08, ‘‘Advanced
Reactor Content of Application
Project, ‘Risk-Informed Technical
Specifications’ ’’
• DANU–ISG–2022–09, ‘‘Advanced
Reactor Content of Application
Project, ‘Risk-Informed, PerformanceBased Fire Protection Program (for
Operations)’ ’’
• RG 1.242, ‘‘Performance-Based
Emergency Preparedness for Small
Modular Reactors, Non-Light-Water
Reactors, and Non-Power Production
or Utilization Facilities’’
• RG 4.7, ‘‘General Site Suitability
Criteria for Nuclear Power Stations’’

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(2) The NRC is issuing for comment
nine draft guidance documents for the
implementation of the proposed
requirements in this rulemaking. The
guidance is available in ADAMS under
the Accession Numbers as indicated
under the ‘‘Availability of Documents’’
section in this document. Comments on
this draft regulatory guidance may be
submitted by the methods outlined in
the ADDRESSES section of this document.
Interested persons may obtain
information and comment submissions
related to the draft guidance by
searching on http://www.regulations.gov
under Docket ID NRC–2019–0062.
• DG–1413, ‘‘Technology-Inclusive
Identification of Licensing Events for
Commercial Nuclear Plants’’
This DG describes an acceptable
approach for identifying licensing
events that can be used to inform the
design basis, licensing basis, and
content of applications for commercial
nuclear plants, including large LWRs
and non-LWRs. It applies to nuclear
power reactor designers, applicants, and
licensees of commercial nuclear plants
applying for permits, licenses,
certifications, and approvals under parts
50, 52, and 53. In this DG, the term
‘‘licensing events’’ is used in a generic
sense to refer to collections of
designated event categories such as, but
not limited to AOOs, DBAs, DBEs, and
postulated accidents. Specifically, this
DG provides an acceptable approach for:
(1) conducting a comprehensive and
systematic search for initiating events;
(2) using a systematic process to
delineate a comprehensive set of event
sequences; (3) grouping initiating events
and event sequences into designated
licensing event categories; and (4)
providing assurance that the set of
licensing events is complete.
• DG–5073, ‘‘Fitness For Duty Programs
for Commercial Nuclear Plants And
Manufacturing Facilities Licensed
Under 10 CFR part 53’’
This DG describes guidance for
applicants under part 53 and licensees
and other entities described in § 26.3(f)
who would elect to or be required to
implement FFD programs for facilities
licensed under part 53. The FFD
program requirements would be
detailed in subpart M of part 26 and
involve, in part, policies, procedures,
drug and alcohol testing, laboratory
requirements, behavioral observation,
MRO responsibilities, fitness
determinations, reporting, and
recordkeeping. The FFD program for
facilities licensed under part 53 subject
to part 26 would also include
requirements for a PMRP and FFD
program change control that licensees or

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other entities must implement to
maintain an effective FFD program.
• DG–5074, ‘‘Access Authorization
Program for Commercial Nuclear
Plants’’
This DG describes a method that the
staff considers acceptable to comply
with requirements in proposed § 73.120,
‘‘Access authorization program for
commercial nuclear plants,’’ related to
an AA program. This document
provides guidance and would be one
NRC-approved method (not the only
method) for meeting regulatory
requirements for part 53. The proposed
language in § 73.120 would provide
flexibility through availability of the use
of an alternate approach, commensurate
with risk and consequence to public
health and safety, for part 53 applicants
who demonstrate in an analysis that the
offsite consequences satisfy the criterion
defined in proposed § 53.860(a)(2)(i).
• DG–5075, ‘‘Establishing Cybersecurity
Programs for Commercial Nuclear
Plants Licensed Under 10 CFR part
53’’
This DG describes an approach the
NRC staff deems acceptable for
complying with the Commission’s
proposed regulations for establishing,
implementing, and maintaining a
cybersecurity program at commercial
nuclear plants that would be licensed
under part 53. This guidance provides
an approach for meeting the
requirements of proposed § 73.110,
‘‘Technology-inclusive requirements for
protection of digital computer and
communication systems and networks.’’
• DG–5076, ‘‘Guidance for Technology
Inclusive Requirements for Physical
Protection of Licensed Activities at
Commercial Nuclear Plants’’
This DG describes methods and
approaches that the NRC staff considers
acceptable for meeting the proposed
physical security requirements of part
53 and § 73.100. The guidance is
intended to provide methods and
considerations for complying with
§ 53.440(f) safety and security design
process considerations, determining
eligibility for meeting the performance
criterion in § 53.860 to relieve the
applicant from the applicable
requirements to defend against
radiological sabotage outlined in § 73.55
or § 73.100, and (if the required analysis
for eligibility is not satisfied) applying
the physical security requirements of
§ 73.100 as an alternative pathway from
§ 73.55 for protection against
radiological sabotage.
• DG–5078, ‘‘Fatigue Management for
Nuclear Power Plant Personnel at

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87011

Commercial Nuclear Plants Licensed
Under 10 CFR part 53’’
This DG describes proposed methods
that the NRC staff considers acceptable
for addressing certain aspects of FFD
programs that would be established at
commercial nuclear facilities licensed
under part 53. This guidance, in
conjunction with the existing RG 5.73,
‘‘Fatigue Management for Nuclear Plant
Personnel,’’ would provide
comprehensive guidance regarding
acceptable methods for the development
and implementation of licensee fatiguemanagement programs.
The NRC is issuing for public
comment the following draft ISG
documents for the implementation of
NRC staff review of applications under
the proposed requirements in this
rulemaking:
• DRO–ISG–2023–01, ‘‘Operator
Licensing Programs’’
This draft ISG provides guidance for
the review of tailored operator licensing
programs that are submitted for review
consistent with the technical
requirements of proposed § 53.730(g).
This guidance primarily addresses the
review of operator licensing
examination processes to facilitate the
ability of reviewers to assess whether a
proposed approach to the testing of
licensed operators and trainees reflects
sound assessment testing practices that
are suitable for the screening of
competent licensed operators.
Additionally, this ISG provides further
review guidance in other areas such as
licensed operator continuing training
and proficiency programs.
• DRO–ISG–2023–02, ‘‘Interim Staff
Guidance Augmenting NUREG–1791,
‘Guidance for Assessing Exemption
Requests from the Nuclear Power
Plant Licensed Operator Staffing
Requirements Specified in 10 CFR
50.54(m),’ for Licensing Commercial
Nuclear Plants under 10 CFR part 53’’
This draft ISG provides guidance for
the review of customized facility
operator staffing plans that are
submitted for review consistent with the
technical requirements of proposed
§ 53.730(f). This ISG is structured as a
companion document to the existing
NUREG–1791 and adapts the existing
HFE-based methodologies of that
document for use in the evaluation of
staffing plans that would be submitted
within the context of part 53 facilities.
Additionally, this ISG provides further
guidance to address other staffingrelated considerations, such as
provisions for engineering expertise.
• DRO–ISG–2023–03, ‘‘Development of
Scalable Human Factors Engineering
Review Plans’’

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Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules

This draft ISG applies to the HFE
review of applications for OLs, COLs,
DCs, and standard design approvals for
commercial nuclear plants submitted
under proposed part 53. The purpose of
this ISG is to facilitate NRC
understanding of an acceptable method
for developing a scalable (i.e.,
application-specific) plan for the review
of these applications for compliance
with applicable HFE requirements. The
ISG describes a process and provides
implementation guidance for the NRC to
tailor HFE review plans to each
application to achieve an effective and
efficient review.
(3) The NRC has identified future
guidance activities that would need to
be completed after the part 53 proposed
rule is published for public comment to
support advanced reactor applications
and NRC reviews. For example, the NRC
recognizes that new guidance would be
needed for the implementation of
provisions in proposed § 53.620(d) and

the associated licensing provisions in
proposed subpart H that would allow
and establish requirements for the
loading of fuel into a manufactured
reactor for subsequent transport to and
use at a commercial nuclear plant that
will operate the facility pursuant to a
COL. The NRC has not yet initiated the
development of guidance documents in
this category but will engage
stakeholders during the development of
these documents to ensure common
prioritization. In addition, the NRC
works with standards development
organizations, advanced reactor
developers, DOE, and other stakeholders
to identify and facilitate new consensus
codes and standards needed for
advanced reactor development. The
NRC will continue its membership and
participation on standards development
committees and working groups to
support standards for advanced reactor
technologies, where appropriate.

XVIII. Public Meeting
The NRC will conduct a public
meeting on this proposed rule for the
purpose of describing the proposed rule
and implementation guidance to the
public and answering questions from
the public on the proposed rule and
implementation guidance.
The NRC will publish a notice of the
public meeting’s location, time, and
agenda on the NRC’s public meeting
website at least 10 calendar days before
the meeting. Stakeholders should
monitor the NRC’s public meeting
website for information about the public
meeting at: https://www.nrc.gov/publicinvolve/public-meetings/index.cfm.
XIX. Availability of Documents
The documents identified in the
following table are available to
interested persons through one or more
of the following methods, as indicated.
ADAMS accession No./Web link/
Federal Register Citation

Document
Proposed Rule Documents
Federal Register Notification, ‘‘Proposed Rule: Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors,’’ October, 2024.
‘‘Draft Environmental Assessment for the Proposed Rule—Risk Informed, Technology-Inclusive
Regulatory Framework for Advanced Reactors,’’ October, 2024.
‘‘Draft Regulatory Analysis for the Proposed Rule: Risk Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors,’’ October, 2024.

ML24095A161.
ML24095A163.
ML24095A166.

Information Collection Documents
Draft Supporting Statement for Information Collection Analysis—10 CFR Part 53 .......................
Draft Supporting Statement for Information Collection Analysis—10 CFR Part 26 .......................
Draft Supporting Statement for Information Collection Analysis—10 CFR Part 50 .......................
Draft Supporting Statement for Information Collection Analysis—10 CFR Part 73 .......................
Draft Supporting Statement for Information Collection Analysis—NRC Form 361S ......................
Draft Supporting Statement for Information Collection Analysis—NRC Form 366 ........................
Draft Supporting Statement for Information Collection Analysis—NRC Form 893 and 894 ..........
Proposed Rule—Part 26 Burden Tables for Risk Informed, Technology-Inclusive Regulatory
Framework for Advanced Reactors.
Proposed Rule—Part 50 Burden Tables for Risk Informed, Technology-Inclusive Regulatory
Framework for Advanced Reactors.
Proposed Rule—Part 53 Burden Tables for Risk Informed, Technology-Inclusive Regulatory
Framework for Advanced Reactors.
Proposed Rule—Part 73 Burden Tables for Risk Informed, Technology-Inclusive Regulatory
Framework for Advanced Reactors.
Draft NRC Form 361S, ‘‘Part 53 Plant Event Notification Worksheet’’ ..........................................
Draft NRC Form 366, ‘‘Licensee Event Report (LER)’’ ..................................................................
Draft NRC Form 366A, ‘‘Licensee Event Report (LER) Continuation Sheet’’ ................................
Draft NRC Form 366B, ‘‘Licensee Event Report (LER) (Failure Continuation)’’ ............................
Draft NRC Form 893, ‘‘10 CFR Part 26, Subpart M, Single FFD Policy Violation Form’’ .............
Draft NRC Form 894, ‘‘10 CFR Part 26, Subpart M, Annual Reporting Form for FFD Performance Information’’.

ML21162A109.
ML23030A400.
ML24220A036.
ML23030A576.
ML24220A034.
ML24220A035.
ML24220A033.
ML24240A008.
ML24220A061.
ML24220A060.
ML24240A009.
ML23032A443.
ML23032A445.
ML23032A447.
ML23032A454.
ML23032A435.
ML23032A439.

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Draft Regulatory Guidance Documents
DG–1413, ‘‘Technology-Inclusive Identification Of Licensing Events For Commercial Nuclear
Plants,’’ October, 2024.
DG–5073, ‘‘Fitness-For-Duty Programs For Commercial Nuclear Plants And Manufacturing Facilities Licensed Under 10 CFR Part 53,’’ October, 2024.
DG–5074, ‘‘Access Authorization Program for Commercial Nuclear Plants,’’ October, 2024 .......
DG–5075, ‘‘Establishing Cybersecurity Programs For Commercial Nuclear Plants Licensed
Under 10 CFR Part 53,’’ October, 2024.
DG–5076, ‘‘Guidance for Technology Inclusive Requirements for Physical Protection of Licensed Activities at Commercial Nuclear Plants,’’ October, 2024.

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ML22200A037.
ML22199A246.
ML22199A257.
ML22203A131.

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87013

ADAMS accession No./Web link/
Federal Register Citation

Document
DG–5078, ‘‘Fatigue Management For Nuclear Power Plant Personnel At Commercial Nuclear
Plants Licensed Under 10 CFR Part 53,’’ October, 2024.

ML22264A109.

Draft ISG Documents
Draft ISG DRO–ISG–2023–01, ‘‘Operator Licensing Programs,’’ October, 2024 ..........................
Draft ISG DRO–ISG–2023–02, ‘‘Interim Staff Guidance Augmenting NUREG–1791, ‘Guidance
for Assessing Exemption Requests from the Nuclear Power Plant Licensed Operator Staffing
Requirements Specified in 10 CFR 50.54(m),’ for Licensing Commercial Nuclear Plants
under 10 CFR Part 53,’’ October, 2024.
Draft ISG DRO–ISG–2023–03, ‘‘Development of Scalable Human Factors Engineering Review
Plans,’’ October, 2024.

ML22266A066.
ML22266A068.

ML22266A072.

Other References
American National Standards Institute/ANS–3.4–2013, ‘‘Medical Certification And Monitoring Of
Personnel Requiring Operator Licenses For Nuclear Power Plants’’.
ASME/ANS RA–S–1.4–2021, ‘‘Probabilistic Risk Assessment Standard for Advanced Non-Light
Water Reactor Nuclear Power Plants’’.
ASCE/SEI 43–19, ‘‘Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities’’.
Federal Register notification—Final policy statement, ‘‘Use of Probabilistic Risk Assessment
Methods in Nuclear Regulatory Activities; Final Policy Statement,’’ dated August 16, 1995.
Federal Register notification—Final rule, ‘‘Fitness-for-Duty Programs,’’ dated June 7, 1989 ......
Federal Register notification—Final rule, ‘‘Fitness for Duty Programs,’’ dated March 31, 2008 ..
Federal Register notification—Final rule, ‘‘Licenses, Certifications, and Approvals for Nuclear
Power Plants,’’ dated August 28, 2007.
Federal Register notification—Final rule, ‘‘Station Blackout,’’ dated June 21, 1988 ....................
Federal Register notification—Final rule, ‘‘Technical Specifications,’’ dated July 19, 1995 .........
Federal Register notification—Guidance, ‘‘Mandatory Guidelines for Federal Workplace Drug
Testing Programs,’’ dated January 23, 2017.
Federal Register notification—Guidance, ‘‘Mandatory Guidelines for Federal Workplace Drug
Testing Programs—Oral/Fluid,’’ dated October 25, 2019.
Federal Register notification—Policy Statement, ‘‘Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants,’’ dated August 8, 1985.
Federal Register notification—Policy Statement, ‘‘Safety Goals for the Operation of Nuclear
Power Plants; Policy Statement; Correction and Republication,’’ dated August 21, 1986.
Federal Register notification—Policy Statement, ‘‘Tribal Policy Statement,’’ dated January 9,
2017.
Federal Register notification—Policy Statement, ‘‘Policy Statement on the Regulation of Advanced Reactors,’’ dated October 14, 2008.
Federal Register notification—Policy Statement, ‘‘Final Safety Culture Policy Statement,’’
dated June 14, 2011.
Federal Register notification—Proposed rule, ‘‘Emergency Preparedness for Small Modular
Reactors and Other New Technologies,’’ dated May 12, 2020.
Federal Register notification—Proposed rule, ‘‘Regulatory Improvements for Production and
Utilization Facilities Transitioning to Decommissioning,’’ dated March 3, 2022.
Federal Register notification—Public meeting, ‘‘Reporting Requirements for Nonemergency
Events at Nuclear Power Plants,’’ dated November 29, 2021.
ICRP, Publication 2 ‘‘Permissible dose for internal radiation,’’ dated 1960 ...................................
ICRP, Publication 26 ‘‘Recommendations of the ICRP,’’ dated 1977 ............................................

lotter on DSK11XQN23PROD with PROPOSALS2

ICRP, Publication 30 ‘‘Limits for Intakes of Radionuclides by Workers,’’ dated 1979 ...................
Letter to Chairman Hanson, NRC, ‘‘Final Letter on Draft 10 CFR Part 53 Rulemaking Language,’’ dated November 22, 2022.
Letter to Chairman Hanson, NRC, ‘‘Fourth Interim Letter on 10 CFR Part 53 Rulemaking Language,’’ dated August 2, 2022.
Letter to Chairman Hanson, NRC, ‘‘Preliminary Proposed Rule Language For 10 CFR Part 53,
Regulation of Advanced Nuclear Reactors, Interim Report,’’ dated May 30, 2021.
Letter to Chairman Hanson, NRC, ‘‘Preliminary Rule Language For 10 CFR Part 53, Subpart F,
‘Requirements for Operations,’ Interim Report,’’ dated February 17, 2022.
Letter to Chairman Rempe, ACRS, ‘‘Response to the Advisory Committee on Reactor Safeguards, ‘Fourth Interim Letter on 10 CFR Part 53 Rulemaking Language,’’’ dated September
30, 2022.
Letter to Chairman Rempe, ACRS, ‘‘Response to the Advisory Committee on Reactor Safeguards Letter on Preliminary Rule Language for 10 CFR Part 53, Subpart F, ‘Requirements
for Operations,’ Interim Report,’’ dated March 30, 2022.
Letter to Chairman Sunseri, ACRS, ‘‘Part 53, Licensing and Regulation of Advanced Nuclear
Reactors,’’ dated November 24, 2020.

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https://webstore.ansi.org/Standards/ANSI/
ansians2013.
https://www.asme.org/codes-standards/findcodes-standards/probabilistic-risk-assessment-standard-for-advanced-non-light-waterreactor-nuclear-power-plants/2021/pdf.
https://doi.org/10.1061/9780784415405.
60 FR 42622.
54 FR 24468.
73 FR 16966.
72 FR 49352.
53 FR 23203.
60 FR 36953, 36955.
82 FR 7920.
84 FR 57554.
50 FR 32138.
51 FR 30028.
82 FR 2402.
73 FR 60612.
76 FR 34773.
85 FR 28436.
87 FR 12254.
86 FR 67669.
https://www.icrp.org/publication.asp?id=icrp%20
publication%202.
https://www.icrp.org/publication.asp?id=
ICRP%20Publication%2026.
https://www.icrp.org/publication.asp?id=
ICRP%20Publication%2030%20(Index).
ML22319A104.
ML22196A292.
ML21140A354.
ML22040A361.
ML22249A073.
ML22063A012.
ML20311A006.

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Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules
ADAMS accession No./Web link/
Federal Register Citation

Document

lotter on DSK11XQN23PROD with PROPOSALS2

Letter to Chairman Svinicki, NRC, ‘‘10 CFR Part 53, Licensing and Regulation of Advanced
Nuclear Reactors,’’ dated October 21, 2020.
Michigan v. EPA, 135 S. Ct. 2699 (2015) ......................................................................................
National Library of Medicine, National Institutes of Health, Workshop Summary, ‘‘The Evolution
of Telehealth: Where Have We Been and Where Are We Going?,’’ dated November 2012.
NEI 18–04, Rev. 1, ‘‘Risk-Informed Performance-Based Technology-Inclusive Guidance for
Non-Light Water Reactors,’’ dated August 2019.
NIA, ‘‘Clarifying ‘Major Portions’ of a Reactor Design in Support of a Standard Design Approval,’’ dated April 2017.
NRC, ‘‘A Regulatory Review Roadmap for Non-Light Water Reactors,’’ dated December 2017 ..
NRC, ‘‘Manufacturing License ML–1 for Production of Up to Eight Floating Nuclear Plants,’’
dated September 30, 1982.
NRC, ‘‘Risk-Informed and Performance-Based Human-System Considerations for Advanced
Reactors,’’ dated March 2021.
NRC Form 890, ‘‘Single Positive Test Form’’ .................................................................................
NRC Form 891, ‘‘Annual Reporting for Drug and Alcohol Tests’’ ..................................................
NRC From 892, ‘‘Annual Fatigue Reporting Form’’ ........................................................................
NUREG–0654/FEMA–REP–1, Revision 2, ‘‘Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants,’’
dated December 2019.
NUREG–0880, ‘‘Safety Goals for Nuclear Power Plant Operation,’’ dated May 1983 ..................
NUREG–1530, Revision 1, ‘‘Reassessment of NRC’s Dollar Per Person-Rem Conversion Factor Policy, Final Report,’’ dated February 2022.
NUREG/BR–0058, Revision 5, ‘‘Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory
Commission,’’ dated April 2017.
NUREG/CR–5884, ‘‘Revised Analyses of Decommissioning for the Reference Pressurized
Water Reactor Power Station,’’ dated November 1995.
NUREG/CR–6187, Volume 1, ‘‘Revised Analyses of Decommissioning for the Reference Boiling Water Reactor Power Station,’’ dated July 1996.
OMB Circular No. A–119, ‘‘Federal Participation in the Development and Use of Voluntary Consensus Standards and in Conformity Assessment Activities,’’ dated February 19, 1998.
PNNL, Technical Letter Report, ‘‘The Use of Electronic Communications to Perform Determinations of Fitness,’’ dated August 2017.
Pre-decisional DG, ‘‘Technology-Inclusive, Risk-Informed, and Performance-Based Methodology for Seismic Design of Commercial Nuclear Plants,’’ dated October 3, 2022.
Research Information Letter 2021–04, ‘‘Feasibility Study on a Potential Consequence-Based
Seismic Design Approach for Nuclear Facilities,’’ dated April 2021.
RG 1.110, Revision 1, ‘‘Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled
Nuclear Power Reactors,’’ dated October 2013.
RG 1.134, Revision 4, ‘‘Medical Assessment Of Licensed Operators Or Applicants For Operator Licenses At Nuclear Power Plants,’’ dated September 2014.
RG 1.174, ‘‘An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions
on Plant-Specific Changes to the Licensing Basis,’’ Revision 3, dated January 2018.
RG 1.208, ‘‘A Performance-Based Approach to Define the Site-Specific Earthquake Ground
Motion,’’ dated March 2007.
RG 1.232, ‘‘Guidance for Developing Principal Design Criteria for Non-Light-Water Reactors,’’
Revision 0, dated April 2018.
RG 1.233, Revision 0, ‘‘Guidance for a Technology-Inclusive, Risk-Informed, and PerformanceBased Methodology to Inform the Licensing Basis and Content of Applications for Licenses,
Certifications, and Approvals for Non-Light-Water Reactors,’’ dated June 2020.
RG 1.247, ‘‘Acceptability of Probabilistic Risk Assessment Results for Non-Light-Water Reactor
Risk-Informed Activities,’’ issued March 2022 for trial use.
RG 5.73, ‘‘Fatigue Management for Nuclear Power Plant Personnel,’’ dated March 20, 2009 ....
RG 5.77, ‘‘Insider Mitigation Program,’’ Revision 1, dated September 08, 2022 ...........................
RG 5.81, ‘‘Target Set Identification and Development for Nuclear Power Reactors,’’ Revision 1,
dated December 2019 (non-public).
SECY–18–0096, ‘‘Functional Containment Performance Criteria For Non-Light-Water-Reactors,’’ dated September 28, 2018.
SECY–19–0117, ‘‘Technology-Inclusive, Risk-Informed, and Performance-Based Methodology
to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and
Approvals for Non-Light-Water Reactors,’’ dated December 2019.
SECY–20–0032, ‘‘Rulemaking Plan on ‘Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors (RIN–3150–AK31; NRC–2019–0062,’’’ dated April 13, 2020.
SECY–20–0070, ‘‘(Redacted) Technical Evaluation of the Security Bounding Time Concept for
Operating Nuclear Power Plants,’’ dated November 8, 2021.
SECY–22–0072, ‘‘Proposed Rule: Alternative Physical Security Requirements for Advanced
Reactors (RIN 3150–AK19),’’ dated August 2, 2022.
SECY–83–293, ‘‘Amendments to 10 CFR 50 Related to Anticipated Transients Without Scram
(ATWS) Events,’’ dated July 19, 1983.
SECY–93–092, ‘‘Issues Pertaining to the Advanced Reactor (PRISM, MHTGR, and PIUS) and
CANDU 3 Designs and their Relationship to Current Regulatory Requirements,’’ dated April
8, 1993.

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ML20295A647.
https://www.ncbi.nlm.nih.gov/books/
NBK207141/.
ML19241A472.
https://www.nuclearinnovationalliance.org/clarifying-major-portions-reactor-design-supportstandard-design-approval.
ML17312B567.
ML20070J215.
ML21069A003.
ML22013B187.
ML22013B240.
ML22013B250.
ML19347D139.
ML071770230.
ML22053A025.
ML17100A480.
ML14008A187.
ML14008A186.
https://obamawhitehouse.archives.gov/omb/circulars_a119_a119fr.
ML18081A607.
ML22276A149.
ML21113A066.
ML13241A052.
ML14189A385.
ML17317A256.
ML070310619.
ML17325A611.
ML20091L698.
ML21235A008.
ML083450028.
ML16342B024.
ML13151A355.
ML18115A157.
ML18311A264 (package).
ML19340A056.
ML20126G265 (package).
ML21334A003 (package).
ML21278A823 (non-public); ML21278A994
(non-public).
ML040210725.

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ADAMS accession No./Web link/
Federal Register Citation

Document
SRM–SECY–10–0121, ‘‘Modifying the Risk-Informed Regulatory Guidance for New Reactors,’’
dated March 2, 2011.
SRM–SECY–17–0100, ‘‘Security Baseline Inspection Program Assessment Results and Recommendations for Program Efficiencies,’’ dated October 8, 2018.
SRM–SECY–20–0032, ‘‘Rulemaking Plan on ‘Risk-Informed, Technology-Inclusive Regulatory
Framework for Advanced Reactors (RIN–3150–AK31; NRC–2019–0062),’’’ dated October 2,
2020.
SRM–SECY–20–0045, ‘‘Population Related Siting Considerations for Advanced Reactors,’’
dated July 30, 2022.
SRM–SECY–98–144, ‘‘Staff Requirements—SECY–98–144—White Paper on Risk-Informed
and Performance-Based Regulations,’’ dated February 24, 1999.
SECY–23–0021, ‘‘Proposed Rule: Risk-Informed, Technology-Inclusive Regulatory Framework
for Advanced Reactors (RIN 3150–AK31),’’ March 1, 2023.
SECY–23–0021, Enclosure 1, ‘‘Draft Federal Register Notification’’ ............................................
SECY–23–0021, Enclosure 2, ‘‘Draft Environmental Assessment for the Proposed Rule—Risk
Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors’’.
SECY–23–0021, Enclosure 3, ‘‘Draft Regulatory Analysis for the Proposed Rule: Risk Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors’’.
SECY–23–0021, Enclosure 4, ‘‘Alternative Approaches Considered for Selected Topics During
the Development of 10 CFR Part 53’’.
SECY–23–0021, Enclosure 5, ‘‘Estimated Resources for The Risk-Informed, Technology-Inclusive Regulatory Framework For Advanced Reactors Rulemaking’’.
Staff Requirements—SECY–23–0021, ‘‘Proposed Rule: Risk-Informed, Technology-Inclusive
Regulatory Framework for Advanced Reactors (RIN 3150–AK31),’’ March 4, 2024.

Throughout the development of this
rule, the NRC may post documents
related to this rule, including public
comments, on the Federal rulemaking
website at https://www.regulations.gov
under Docket ID NRC–2019–0062. The
Federal rulemaking website allows you
to receive alerts when changes or
additions occur in a docket folder. To
subscribe: (1) Navigate to the docket
folder (NRC–2019–0062–0012); (2) click
the ‘‘Sign up for Email Alerts’’ link; and
(3) enter your email address and select
how frequently you would like to
receive emails (daily, weekly, or
monthly).
List of Subjects
10 CFR Part 1

lotter on DSK11XQN23PROD with PROPOSALS2

10 CFR Part 2
Administrative practice and
procedure, Antitrust, Byproduct
material, Classified information,
Confidential business information,
Freedom of information, Environmental
protection, Hazardous waste, Nuclear
energy, Nuclear materials, Nuclear
power plants and reactors, Penalties,
Reporting and recordkeeping
requirements, Sex discrimination,
Source material, Special nuclear
material, Waste treatment and disposal.

ML22194A885.
ML003753593.
ML21162A095.
ML21162A102.
ML21162A104.
ML21165A112.
ML22244A001.
ML22304A099 (non-public).
ML24064A047 (package).

10 CFR Part 26

Hazardous materials transportation,
Investigations, Nuclear energy, Nuclear
materials, Penalties, Reporting and
recordkeeping requirements, Security
measures, Special nuclear material.

Administrative practice and
procedure, Alcohol abuse, Alcohol
testing, Appeals, Drug abuse, Drug
testing, Employee assistance programs,
Fitness for duty, Management actions,
Nuclear power plants and reactors,
Privacy, Protection of information,
Radiation protection, Reporting and
recordkeeping requirements.

10 CFR Part 19
Criminal penalties, Environmental
protection, Nuclear Energy, Nuclear
materials, Nuclear power plants and
reactors, Occupational safety and
health, Penalties, Radiation protection,
Reporting and recordkeeping
requirements, Sex discrimination.
Byproduct material, Criminal
penalties, Hazardous waste, Licensed
material, Nuclear energy, Nuclear
materials, Nuclear power plants and
reactors, Occupational safety and
health, Packaging and containers,
Penalties, Radiation protection,
Reporting and recordkeeping
requirements, Source material, Special
nuclear material, Waste treatment and
disposal.
10 CFR Part 21
Nuclear power plants and reactors,
Penalties, Radiation protection,
Reporting and recordkeeping
requirements.

Administrative practice and
procedure, Classified information,

Classified information, Criminal
penalties, Investigations, Penalties,

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10 CFR Part 11

10 CFR Part 25

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ML18283A072.

Reporting and recordkeeping
requirements, Security measures.

10 CFR Part 10

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ML110610166.

Government employees, Security
measures.

10 CFR Part 20

Flags, Organization and functions
(Government Agencies), Seals and
insignia.

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10 CFR Part 30
Byproduct material, Criminal
penalties, Government contracts,
Intergovernmental relations, Isotopes,
Nuclear energy, Nuclear materials,
Penalties, Radiation protection,
Reporting and recordkeeping
requirements, Whistleblowing.
10 CFR Part 40
Criminal penalties, Exports,
Government contracts, Hazardous
materials transportation, Hazardous
waste, Nuclear energy, Nuclear
materials, Penalties, Reporting and
recordkeeping requirements, Source
material, Uranium, Whistleblowing.
10 CFR Part 50
Administrative practice and
procedure, Antitrust, Backfitting,
Classified information, Criminal
penalties, Education, Emergency
planning, Fire prevention, Fire
protection, Intergovernmental relations,
Nuclear power plants and reactors,
Penalties, Radiation protection, Reactor
siting criteria, Reporting and

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10 CFR Part 75
Criminal penalties, Intergovernmental
relations, Nuclear energy, Nuclear
materials, Nuclear power plants and
reactors, Penalties, Reporting and
recordkeeping requirements, Security
measures, Treaties.

recordkeeping requirements,
Whistleblowing.
10 CFR Part 51
Administrative practice and
procedure, Environmental impact
statements, Hazardous waste, Nuclear
energy, Nuclear materials, Nuclear
power plants and reactors, Reporting
and recordkeeping requirements.
10 CFR Part 53
Administrative practice and
procedure, Antitrust, Backfitting,
Construction permit, Combined license,
Classified information, Criminal
penalties, Early site permit, Emergency
planning, Fees, Fire prevention, Fire
protection, Inspection,
Intergovernmental relations, Limited
work authorization, Manufacturing
license, Nuclear power plants and
reactors, Operating license, Penalties,
Prototype, Radiation protection, Reactor
siting criteria, Reporting and
recordkeeping requirements, Standard
design, Standard design certification,
Training programs.
10 CFR Part 70
Classified information, Criminal
penalties, Emergency medical services,
Hazardous materials transportation,
Material control and accounting,
Nuclear energy, Nuclear materials,
Packaging and containers, Penalties,
Radiation protection, Reporting and
recordkeeping requirements, Scientific
equipment, Security measures, Special
nuclear material, Whistleblowing.
10 CFR Part 72
Administrative practice and
procedure, Hazardous waste, Indians,
Intergovernmental relations, Nuclear
energy, Penalties, Radiation protection,
Reporting and recordkeeping
requirements, Security measures, Spent
fuel, Whistleblowing.
10 CFR Part 73
Criminal penalties, Exports,
Hazardous materials transportation,
Imports, Nuclear energy, Nuclear
materials, Nuclear power plants and
reactors, Penalties, Reporting and
recordkeeping requirements, Security
measures.
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10 CFR Part 74
Accounting, Criminal penalties,
Hazardous materials transportation,
Material control and accounting,
Nuclear energy, Nuclear materials,
Packaging and containers, Penalties,
Radiation protection, Reporting and
recordkeeping requirements, Scientific
equipment, Special nuclear material.

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10 CFR Part 95
Classified information, Criminal
penalties, Penalties, Reporting and
recordkeeping requirements, Security
measures.
10 CFR Part 140
Criminal penalties, Extraordinary
nuclear occurrence, Insurance,
Intergovernmental relations, Nuclear
materials, Nuclear power plants and
reactors, Penalties, Reporting and
recordkeeping requirements.
10 CFR Part 150
Criminal penalties, Hazardous
materials transportation,
Intergovernmental relations, Nuclear
energy, Nuclear materials, Penalties,
Reporting and recordkeeping
requirements, Security measures,
Source material, Special nuclear
material.

Reorganization Act of 1974, secs. 201, 203,
204, 205, 209 (42 U.S.C. 5841, 5843, 5844,
5845, 5849); Administrative Procedure Act
(5 U.S.C. 552, 553); Reorganization Plan No.
1 of 1980, 5 U.S.C. Appendix (Reorganization
Plans).
§ 1.43

PART 2—AGENCY RULES OF
PRACTICE AND PROCEDURE
3. The authority citation for part 2
continues to read as follows:

■

Authority: Atomic Energy Act of 1954,
secs. 29, 53, 62, 63, 81, 102, 103, 104, 105,
161, 181, 182, 183, 184, 186, 189, 191, 234
(42 U.S.C. 2039, 2073, 2092, 2093, 2111,
2132, 2133, 2134, 2135, 2201, 2231, 2232,
2233, 2234, 2236, 2239, 2241, 2282); Energy
Reorganization Act of 1974, secs. 201, 206
(42 U.S.C. 5841, 5846); Nuclear Waste Policy
Act of 1982, secs. 114(f), 134, 135, 141
(42 U.S.C. 10134(f), 10154, 10155, 10161);
Administrative Procedure Act (5 U.S.C. 552,
553, 554, 557, 558); National Environmental
Policy Act of 1969 (42 U.S.C. 4332); 44 U.S.C.
3504 note. Section 2.205(j) also issued under
28 U.S.C. 2461 note.

10 CFR Part 170
Byproduct material, Import and
export licenses, Intergovernmental
relations, Non-payment penalties,
Nuclear energy, Nuclear materials,
Nuclear power plants and reactors,
Source material, Special nuclear
material.

§ 2.1

10 CFR Part 171
Annual charges, Approvals,
Byproduct material, Holders of
certificates, Intergovernmental relations,
Nonpayment penalties, Nuclear
materials, Nuclear power plants and
reactors, Registrations, Source material,
Special nuclear material.
For the reasons set out in the
preamble and under the authority of the
Atomic Energy Act of 1954, as amended;
the Energy Reorganization Act of 1974,
as amended; and 5 U.S.C. 552 and 553,
the NRC is proposing the following
amendments to 10 CFR parts 1, 2, 10,
11, 19, 20, 21, 25, 26, 30, 40, 50, 51, 70,
72, 73, 74, 75, 95, 140, 150, 170, and 171
and adding 10 CFR part 53:

*

PART 1—STATEMENT OF
ORGANIZATION AND GENERAL
INFORMATION
1. The authority citation for part 1
continues to read as follows:

■

Authority: Atomic Energy Act of 1954,
secs. 23, 25, 29, 161, 191 (42 U.S.C. 2033,
2035, 2039, 2201, 2241); Energy

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[Amended]

2. In § 1.43, in paragraph (a)(2) remove
the cross reference ‘‘10 CFR parts 50, 52,
and 54’’ and add in its place the cross
reference ‘‘10 CFR parts 50, 52, 53, and
54’’.

■

[Amended]

4. In § 2.1, in paragraph (e) remove the
phrase ‘‘part 52’’ and add in its place
the phrase ‘‘part 52 or part 53’’.
■ 5. In § 2.4, revise the definitions for
‘‘Contested proceeding’’ and ‘‘Facility’’
to read as follows:
■

§ 2.4

Definitions.

*
*
*
*
Contested proceeding means—
(1) A proceeding in which there is a
controversy between the NRC staff and
the applicant for a license or permit
concerning the issuance of the license or
permit or any of the terms or conditions
thereof;
(2) A proceeding in which the NRC is
imposing a civil penalty or other
enforcement action, and the subject of
the civil penalty or enforcement action
is an applicant for or holder of a license
or permit, or is or was an applicant for
or holder of a license or permit, or is or
was an applicant for a standard design
certification under part 52 or part 53 of
this chapter; and
(3) A proceeding in which a petition
for leave to intervene in opposition to
an application for a license or permit
has been granted or is pending before
the Commission.
*
*
*
*
*
Facility means production facility or a
utilization facility as defined in §§ 50.2
and 53.020 of this chapter.
*
*
*
*
*

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Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules
§ 2.100

[Amended]

6. In § 2.100, remove the phrase
‘‘subpart E of part 52’’ and add in its
place the phrase ‘‘subpart E of part 52
or subpart H of part 53’’.
■ 7. In § 2.101, revise paragraphs
(a)(3)(i), (a)(5), (a)(9) introductory text
and paragraph (a)(9)(i) to read as
follows:
■

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§ 2.101

Filing of application.

(a) * * *
(3) * * *
(i) Submit to the Director, Office of
Nuclear Reactor Regulation, or Director,
Office of Nuclear Material Safety and
Safeguards, as appropriate, such
additional copies as the regulations in
part 50, subpart A of part 51, and part
53 of this chapter require;
*
*
*
*
*
(5) An applicant for a construction
permit under parts 50 or 53 of this
chapter or a combined license under
parts 52 or 53 of this chapter for a
production or utilization facility which
is subject to § 51.20(b) of this chapter,
and is of the type specified in
§ 50.21(b)(2) or (b)(3); or § 50.22; or part
53, as applicable, of this chapter, or is
a testing facility, may submit the
information required of applicants by
parts 50, 52, or 53 of this chapter in two
parts. One part shall be accompanied by
the information required by § 50.30(f) of
this chapter, § 52.80(b) of this chapter,
or § 53.1100(f) of this chapter, as
applicable. The other part shall include
any information required by § 50.34(a)
and, if applicable, § 50.34a of this
chapter; or §§ 52.79 and 52.80(a) of this
chapter; or §§ 53.1109, 53.1306,
53.1309, and 53.1312 of this chapter; or
§§ 53.1109, 53.1413, 53.1416, and
53.1419 of this chapter, as applicable.
One part may precede or follow other
parts by no longer than 6 months. If it
is determined that either of the parts as
described above is incomplete and not
acceptable for processing, the Director,
Office of Nuclear Reactor Regulation, or
Director, Office of Nuclear Material
Safety and Safeguards, as appropriate,
will inform the applicant of this
determination and the respects in which
the document is deficient. Such a
determination of completeness will
generally be made within a period of 30
days. Whichever part is filed first shall
also include the fee required by
§ 50.30(e) or § 53.1100(e) and § 170.21 of
this chapter and the information
required by §§ 50.33, 50.34(a)(1), and
52.79(a)(1) of this chapter; or
§§ 53.1109, 53.1309, and 53.1416 of this
chapter, as applicable, and § 50.37 or
§ 53.1115, as applicable, of this chapter.
The Director, Office of Nuclear Reactor
Regulation, or Director, Office of

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Nuclear Material Safety and Safeguards,
as appropriate, will accept for docketing
an application for a construction permit
under part 50 or part 53 of this chapter
or a combined license under parts 52 or
53 of this chapter for a production or
utilization facility which is subject to
§ 51.20(b) of this chapter, and is of the
type specified in § 50.21(b)(2) or (b)(3),
or § 50.22, or part 53, as applicable, of
this chapter or is a testing facility where
one part of the application as described
above is complete and conforms to the
requirements of part 50 of this chapter.
The additional parts will be docketed
upon a determination by the Director,
Office of Nuclear Reactor Regulation, or
Director, Office of Nuclear Material
Safety and Safeguards, as appropriate,
that it is complete.
*
*
*
*
*
(9) An applicant for a construction
permit for a utilization facility which is
subject to § 51.20(b) of this chapter and
is of the type specified in § 50.21(b)(2)
or (b)(3), or § 50.22, or part 53 of this
chapter, an applicant for or holder of an
early site permit under part 52 or part
53 of this chapter, or an applicant for a
combined license under parts 52 or 53
of this chapter, who seeks to conduct
the activities authorized under
§ 50.10(d) or § 53.1130 of this chapter
may submit a complete application
under paragraphs (a)(1) through (a)(4) of
this section which includes the
information required by § 50.10(d) or
§ 53.1130 of this chapter. Alternatively,
the applicant (other than an applicant
for or holder of an early site permit) may
submit its application in two parts:
(i) Part one must include the
information required by § 50.33(a)
through (f) or § 53.1109(a) through (e)
and § 53.1306 of this chapter, and the
information required by § 50.10(d)(2)
and (d)(3) or § 53.1130(a)(2) and (a)(3) of
this chapter, as applicable.
*
*
*
*
*
■ 8. In § 2.104, revise paragraph (a) to
read as follows:
§ 2.104

Notice of hearing.

(a) In the case of an application on
which a hearing is required by the Act
or this chapter, or in which the
Commission finds that a hearing is
required in the public interest, the
Secretary will issue a notice of hearing
to be published in the Federal Register.
The notice must be published at least 15
days, and in the case of an application
concerning a limited work
authorization, construction permit, early
site permit, or combined license for a
facility of the type described in
§ 50.21(b) or 50.22, or subpart H of part
53 of this chapter, as applicable, or a

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87017

testing facility, at least 30 days, before
the date set for hearing in the notice.1
In addition, in the case of an application
for a limited work authorization,
construction permit, early site permit, or
combined license for a facility of the
type described in § 50.22 or subpart H
of part 53 of this chapter, as applicable,
or a testing facility, the notice must be
issued as soon as practicable after the
NRC has docketed the application. If the
Commission decides, under
§ 2.101(a)(2), to determine the
acceptability of the application based on
its technical adequacy as well as
completeness, the notice must be issued
as soon as practicable after the
application has been tendered.
*
*
*
*
*
1 If the notice of hearing concerning an
application for a limited work authorization,
construction permit, early site permit, or
combined license for a facility of the type
described in § 50.21(b) or § 50.22, or subpart
H of part 53 of this chapter, as applicable, or
a testing facility, does not specify the time
and place of initial hearing, a subsequent
notice will be published in the Federal
Register which will provide at least 30-day
notice of the time and place of that hearing.
After this notice is given, the presiding
officer may reschedule the commencement of
the initial hearing for a later date or
reconvene a recessed hearing without again
providing at least 30-day notice.

9. In § 2.105, revise paragraph (a)
introductory text and paragraphs (a)(4),
(a)(10), (a)(12), (a)(13), (b)(3)
introductory text, (b)(3)(i), (ii), and (iv)
to read as follows:

■

§ 2.105

Notice of proposed action.

(a) If a hearing is not required by the
Act or this chapter, and if the
Commission has not found that a
hearing is in the public interest, it will,
before acting thereon, publish in the
Federal Register, as applicable, or on
the NRC’s website, http://www.nrc.gov,
or both, at the Commission’s discretion,
either a notice of intended operation
under § 52.103(a) or § 53.1452(a) of this
chapter, as applicable, and a proposed
finding that inspections, tests, analyses,
and acceptance criteria for a combined
license under subpart C of part 52 or
under subpart H of part 53 of this
chapter, have been or will be met, or a
notice of proposed action with respect
to an application for:
*
*
*
*
*
(4) An amendment to an operating
license, combined license, or
manufacturing license for a facility
licensed under § 50.21(b) or § 50.22 or
under subpart H of part 53 of this
chapter, as applicable, or for a testing
facility, as follows:
(i) If the Commission determines
under § 50.58 or § 53.1515 of this

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Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules

chapter that the amendment involves no
significant hazards consideration,
though it will provide notice of
opportunity for a hearing pursuant to
this section, it may make the
amendment immediately effective and
grant a hearing thereafter; or
(ii) If the Commission determines
under §§ 50.58 and 50.91 or § 53.1515 of
this chapter, as applicable, that an
emergency situation exists or that
exigent circumstances exist and that the
amendment involves no significant
hazards consideration, it will provide
notice of opportunity for a hearing
pursuant to § 2.106 (if a hearing is
requested, it will be held after issuance
of the amendment);
*
*
*
*
*
(10) In the case of an application for
an operating license for a facility of a
type described in § 50.21(b) or § 50.22,
or part 53 of this chapter or a testing
facility, a notice of opportunity for
hearing shall be issued as soon as
practicable after the application has
been docketed; or
*
*
*
*
*
(12) An amendment to an early site
permit issued under subpart A of part
52, or under subpart H of part 53 of this
chapter, as follows:
(i) If the early site permit does not
provide authority to conduct the
activities allowed under § 50.10(e)(1) or
§ 53.1130(b)(1) of this chapter, the
amendment will involve no significant
hazards consideration, and though the
NRC will provide notice of opportunity
for a hearing under this section, it may
make the amendment immediately
effective and grant a hearing thereafter;
and
(ii) If the early site permit provides
authority to conduct the activities
allowed under § 50.10(e)(1) or
§ 53.1130(b)(1) of this chapter and the
Commission determines under §§ 50.58
and 50.91 or § 53.1515 of this chapter
that an emergency situation exists or
that exigent circumstances exist and
that the amendment involves no
significant hazards consideration, it will
provide notice of opportunity for a
hearing under § 2.106 of this chapter (if
a hearing is requested, which will be
held after issuance of the amendment).
(13) A manufacturing license under
subpart F of part 52 or subpart H of part
53 of this chapter.
(b) * * *
(3) For a notice of intended operation
under § 52.103(a) or § 53.1452(a) of this
chapter, the following information:
(i) The identification of the NRC
action as making the finding required
under § 52.103(g) or § 53.1452(g) of this
chapter;

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(ii) The manner in which the licensee
notifications under § 52.99(c) or
§ 53.1449(c), of this chapter which are
required to be made available by
§ 52.99(e)(2) or § 53.1449(e)(2), of this
chapter may be obtained and examined;
*
*
*
*
*
(iv) Any conditions, limitations, or
restrictions to be placed on the license
in connection with the finding under
§ 52.103(g) or § 53.1452(g) of this
chapter, and the expiration date or
circumstances (if any) under which the
conditions, limitations or restrictions
will no longer apply.
*
*
*
*
*
■ 10. In § 2.106, revise paragraphs (a)(2),
(a)(3), and (b)(2) introductory text to
read as follows:
§ 2.106

Notice of issuance.

(a) * * *
(2) An amendment of a license for a
facility of the type described in
§ 50.21(b) or § 50.22, or part 53 of this
chapter, as applicable, or a testing
facility, whether or not a notice of
proposed action has been previously
published; and
(3) The finding under § 52.103(g) or
§ 53.1452(g) of this chapter.
(b) * * *
(2) In the case of a finding under
§ 52.103(g) or § 53.1452(g) of this
chapter:
*
*
*
*
*
■ 11. In § 2.109, revise paragraphs (b),
(c), and (d) to read as follows:
§ 2.109 Effect of timely renewal
application.

*

*
*
*
*
(b) If the licensee of a nuclear power
plant licensed under § 50.21(b) or
§ 50.22 or under subpart H of part 53 of
this chapter files a sufficient application
for renewal of either an operating
license or a combined license at least 5
years before the expiration of the
existing license, the existing license will
not be deemed to have expired until the
application has been finally determined.
(c) If the holder of an early site permit
licensed under subpart A of part 52 or
under subpart H of part 53 of this
chapter, as applicable, files a sufficient
application for renewal under § 52.29 or
§ 53.1173 of this chapter, as applicable,
at least 12 months before the expiration
of the existing early site permit, the
existing permit will not be deemed to
have expired until the application has
been finally determined.
(d) If the licensee of a manufacturing
license under subpart F of part 52, or
under subpart H of part 53 of this
chapter files a sufficient application for
renewal under § 52.177 or § 53.1295 of
this chapter at least 12 months before

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the expiration of the existing license,
the existing license will not be deemed
to have expired until the application has
been finally determined.
*
*
*
*
*
■ 12. In § 2.110, revise paragraphs (a)(1)
and (b) to read as follows:
§ 2.110 Filing and administrative action on
submittals for standard design approval or
early review of site suitability issues.

(a)(1) A submittal for a standard
design approval under subpart E of part
52 or under subpart H of part 53 of this
chapter shall be subject to §§ 2.101(a)
and 2.390 to the same extent as if it were
an application for a permit or license.
*
*
*
*
*
(b) Upon initiation of review by the
NRC staff of a submittal for an early
review of site suitability issues under
appendix Q to part 50 of this chapter,
or for a standard design approval under
subpart E of part 52 or under subpart H
of part 53 of this chapter, the Director,
Office of Nuclear Reactor Regulation,
shall publish in the Federal Register a
notice of receipt of the submittal,
inviting comments from interested
persons within 60 days of publication or
other time as may be specified, for
consideration by the NRC staff and
ACRS in their review.
*
*
*
*
*
■ 13. In § 2.202, revise paragraph (e) to
read as follows:
§ 2.202

Orders.

*

*
*
*
*
(e)(1) If the order involves the
modification of a part 50 or a part 53
license and is a backfit, the
requirements of § 50.109 or § 53.1590 of
this chapter, as applicable, shall be
followed, unless the licensee has
consented to the action required.
(2) If the order involves the
modification of combined license under
subpart C of part 52, or subpart H of part
53 of this chapter, the requirements of
§ 52.98 or § 53.1443 of this chapter, as
applicable, shall be followed unless the
licensee has consented to the action
required.
(3) If the order involves a change to
an early site permit under subpart A of
part 52 or under subpart H of part 53 of
this chapter, the requirements of § 52.39
or § 53.1188 of this chapter, as
applicable, must be followed, unless the
applicant or licensee has consented to
the action required.
(4) If the order involves a change to
a standard design certification rule
referenced by that plant’s application,
the requirements, if any, in the
referenced design certification rule with
respect to changes must be followed, or,
in the absence of these requirements,

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Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules
the requirements of § 52.63 or § 53.1263
of this chapter, as applicable, must be
followed, unless the applicant or
licensee has consented to follow the
action required.
(5) If the order involves a change to
a standard design approval referenced
by that plant’s application, the
requirements of § 52.145 or § 53.1221 of
this chapter, as applicable, must be
followed unless the applicant or
licensee has consented to follow the
action required.
(6) If the order involves a
modification of a manufacturing license
under subpart F of part 52 or under
subpart H of part 53 of this chapter, the
requirements of § 52.171 or § 53.1288 of
this chapter, as applicable, must be
followed, unless the applicant or
licensee has consented to the action
required.
■ 14. In § 2.309, revise paragraphs (a),
(f)(1)(i), (f)(1)(vi) and (vii), (g), (h)(2),
(i)(2), (j) to read as follows:

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§ 2.309 Hearing requests, petitions to
intervene, requirements for standing, and
contentions.

(a) General requirements. Any person
whose interest may be affected by a
proceeding and who desires to
participate as a party must file a written
request for hearing and a specification
of the contentions which the person
seeks to have litigated in the hearing. In
a proceeding under § 52.103 or
§ 53.1452 of this chapter, as applicable,
the Commission, acting as the presiding
officer, will grant the request if it
determines that the requestor has
standing under the provisions of
paragraph (d) of this section and has
proposed at least one admissible
contention that meets the requirements
of paragraph (f) of this section. For all
other proceedings, except as provided in
paragraph (e) of this section, the
Commission, presiding officer, or the
Atomic Safety and Licensing Board
designated to rule on the request for
hearing and/or petition for leave to
intervene, will grant the request/petition
if it determines that the requestor/
petitioner has standing under the
provisions of paragraph (d) of this
section and has proposed at least one
admissible contention that meets the
requirements of paragraph (f) of this
section. In ruling on the request for
hearing/petition to intervene submitted
by petitioners seeking to intervene in
the proceeding on the HLW repository,
the Commission, the presiding officer,
or the Atomic Safety and Licensing
Board shall also consider any failure of
the petitioner to participate as a
potential party in the pre-license
application phase under subpart J of this

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part in addition to the factors in
paragraph (d) of this section. If a request
for hearing or petition to intervene is
filed in response to any notice of
hearing or opportunity for hearing, the
applicant/licensee shall be deemed to be
a party.
*
*
*
*
*
(f) * * *
(1) * * *
(i) Provide a specific statement of the
issue of law or fact to be raised or
controverted, provided further, that the
issue of law or fact to be raised in a
request for hearing under § 52.103(b) or
§ 53.1452(b) of this chapter, as
applicable, must be directed at
demonstrating that one or more of the
acceptance criteria in the combined
license have not been, or will not be
met, and that the specific operational
consequences of nonconformance
would be contrary to providing
reasonable assurance of adequate
protection of the public health and
safety;
*
*
*
*
*
(vi) In a proceeding other than one
under § 52.103 or § 53.1452 of this
chapter provide sufficient information
to show that a genuine dispute exists
with the applicant/licensee on a
material issue of law or fact. This
information must include references to
specific portions of the application
(including the applicant’s
environmental report and safety report)
that the petitioner disputes and the
supporting reasons for each dispute, or,
if the petitioner believes that the
application fails to contain information
on a relevant matter as required by law,
the identification of each failure and the
supporting reasons for the petitioner’s
belief; and
(vii) In a proceeding under § 52.103(b)
or § 53.1452(b) of this chapter, as
applicable, the information must be
sufficient, and include supporting
information showing, prima facie, that
one or more of the acceptance criteria in
the combined license have not been, or
will not be met, and that the specific
operational consequences of
nonconformance would be contrary to
providing reasonable assurance of
adequate protection of the public health
and safety. This information must
include the specific portion of the report
required by § 52.99(c) or § 53.1449(c) of
this chapter, as applicable, which the
requestor believes is inaccurate,
incorrect, and/or incomplete (i.e., fails
to contain the necessary information
required by § 52.99(c) or § 53.1449(c) of
this chapter, as applicable). If the
requestor identifies a specific portion of
the report under § 52.99(c) or

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87019

§ 53.1449(c) of this chapter, as
applicable, as incomplete and the
requestor contends that the incomplete
portion prevents the requestor from
making the necessary prima facie
showing, then the requestor must
explain why this deficiency prevents
the requestor from making the prima
facie showing.
*
*
*
*
*
(g) Selection of hearing procedures. A
request for hearing and/or petition for
leave to intervene may, except in a
proceeding under § 52.103 or § 53.1452
of this chapter, as applicable, also
address the selection of hearing
procedures, taking into account the
provisions of § 2.310. If a request/
petition relies upon § 2.310(d), the
request/petition must demonstrate, by
reference to the contention and the
bases provided and the specific
procedures in subpart G of this part, that
resolution of the contention necessitates
resolution of material issues of fact
which may be best determined through
the use of the identified procedures.
(h) * * *
(2) If the proceeding pertains to a
production or utilization facility (as
defined in § 50.2 or § 53.020 of this
chapter) located within the boundaries
of the State, local governmental body, or
Federally-recognized Indian Tribe
seeking to participate as a party, no
further demonstration of standing is
required. If the production or utilization
facility is not located within the
boundaries of the State, local
governmental body, or Federallyrecognized Indian Tribe seeking to
participate as a party, the State, local
governmental body, or Federallyrecognized Indian Tribe also must
demonstrate standing.
*
*
*
*
*
(i) * * *
(2) Except in a proceeding under
§ 52.103 or § 53.1452 of this chapter, as
applicable, the participant who filed the
hearing request, intervention petition, or
motion for leave to file new or amended
contentions after the deadline may file
a reply to any answer. The reply must
be filed within 7 days after service of
that answer.
*
*
*
*
*
(j) Decision on request/petition. (1) In
all proceedings other than a proceeding
under § 52.103 or § 53.1452 of this
chapter, as applicable, the presiding
officer shall issue a decision on each
request for hearing or petition to
intervene within 45 days of the
conclusion of the initial pre-hearing
conference or, if no pre-hearing
conference is conducted, within 45 days
after the filing of answers and replies

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under paragraph (i) of this section. With
respect to a request to admit amended
or new contentions, the presiding officer
shall issue a decision on each such
request within 45 days of the conclusion
of any pre-hearing conference that may
be conducted regarding the proposed
amended or new contentions or, if no
pre-hearing conference is conducted,
within 45 days after the filing of
answers and replies, if any. In the event
the presiding officer cannot issue a
decision within 45 days, the presiding
officer shall issue a notice advising the
Commission and the parties, and the
notice shall include the expected date of
when the decision will issue.
(2) The Commission, acting as the
presiding officer, shall expeditiously
grant or deny the request for hearing in
a proceeding under § 52.103 or
§ 53.1452 of this chapter, as applicable.
The Commission’s decision may not be
the subject of any appeal under § 2.311.
■ 15. Amend § 2.310 by:
■ a. In paragraphs (a) and (h)
introductory text, removing the crossreference ‘‘parts 30, 32 through 36, 39,
40, 50, 52, 54, 55, 61, 70 and 72 of this
chapter’’ and adding, in its place, the
cross reference ‘‘parts 30, 32 through 36,
39, 40, 50, 52, 53, 54, 55, 61, 70 and 72
of this chapter’’; and
■ b. Revising paragraphs (i) and (j).
The revisions read as follows.

involving a construction permit or
operating license for a facility of a type
described in §§ 50.21(b) or 50.22 or part
53 of this chapter must be held within
sixty (60) days after discovery has been
completed or any other time specified
by the Commission or the presiding
officer.
*
*
*
*
*
■ 17. In § 2.339, revise paragraph (d) to
read as follows:

§ 2.310

*

Selection of hearing procedures.

*

*
*
*
*
(i) In design certification rulemaking
proceedings under part 52 or part 53 of
this chapter, any informal hearing held
under § 52.51 or § 53.1242 of this
chapter, as applicable, must be
conducted under the procedures of
subpart O of this part.
(j) Proceedings on a Commission
finding under § 52.103(c) and (g) or
§ 53.1452(c) and (g) of this chapter, as
applicable, shall be conducted in
accordance with the procedures
designated by the Commission in each
proceeding.
*
*
*
*
*
■ 16. In § 2.329, revise paragraph (a) to
read as follows:

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§ 2.329

Prehearing conference.

(a) Necessity for prehearing
conference; timing. The Commission or
the presiding officer may, and in the
case of a proceeding on an application
for a construction permit or an operating
license for a facility of a type described
in §§ 50.21(b) or 50.22, or part 53 of this
chapter, or a testing facility, must direct
the parties or their counsel to appear at
a specified time and place for a
conference or conferences before trial. A
prehearing conference in a proceeding

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18:06 Oct 30, 2024

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§ 2.339 Expedited decision-making
procedure.

*

*
*
*
*
(d) The provisions of this section do
not apply to an initial decision directing
the issuance of a limited work
authorization under 10 CFR 50.10 or
10 CFR 53.1130; an early site permit
under subpart A of part 52 or under
subpart H of part 53 of this chapter; a
construction permit or construction
authorization under part 50 or part 53
of this chapter; a combined license
under subpart C of part 52 or under
subpart H of part 53 of this chapter; or
a manufacturing license under subpart F
of part 52 or under subpart H of part 53.
■ 18. In § 2.340, revise paragraphs (b),
(c), (d), (f), (i), and (j) to read as follows:
§ 2.340 Initial decision in certain contested
proceedings; immediate effectiveness of
initial decisions; issuance of authorizations,
permits and licenses.

*
*
*
*
(b) Initial decision—combined license
under 10 CFR parts 52 or 53. (1) Matters
in controversy; presiding officer
consideration of matters not put in
controversy by parties. In any initial
decision in a contested proceeding on
an application for a combined license
under parts 52 or 53 of this chapter
(including an amendment to or renewal
of combined license), the presiding
officer shall make findings of fact and
conclusions of law on the matters put
into controversy by the parties and any
matter designated by the Commission to
be decided by the presiding officer. The
presiding officer shall also make
findings of fact and conclusions of law
on any matter not put into controversy
by the parties, but only to the extent that
the presiding officer determines that a
serious safety, environmental, or
common defense and security matter
exists, and the Commission approves of
an examination of and decision on the
matter upon its referral by the presiding
officer under, inter alia, the provisions
of §§ 2.323 and 2.341.
(2) Presiding officer initial decision
and issuance of permit or license.
(i) In a contested proceeding for the
initial issuance or renewal of a
combined license under parts 52 or 53

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Fmt 4701

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of this chapter, or the amendment of a
combined license where the NRC has
not made a determination of no
significant hazards consideration, the
Commission or the Director, Office of
Nuclear Reactor Regulation, as
appropriate after making the requisite
findings, shall issue, deny, or
appropriately condition the permit or
license in accordance with the presiding
officer’s initial decision once that
decision becomes effective.
(ii) In a contested proceeding for the
amendment of a combined license
under parts 52 or 53 of this chapter
where the NRC has made a
determination of no significant hazards
consideration, the Commission or the
Director, Office of Nuclear Reactor
Regulation, as appropriate (appropriate
official), after making the requisite
findings and complying with any
applicable provisions of § 2.1202(a) or
§ 2.1403(a), may issue the amendment
before the presiding officer’s initial
decision becomes effective. Once the
presiding officer’s initial decision
becomes effective, the appropriate
official shall take action with respect to
that amendment in accordance with the
initial decision. If the presiding officer’s
initial decision becomes effective before
the appropriate official issues the
amendment, then the appropriate
official, after making the requisite
findings, shall issue, deny, or
appropriately condition the amendment
in accordance with the presiding
officer’s initial decision.
(c) Initial decision on findings under
10 CFR 52.103 or 10 CFR 53.1452 with
respect to acceptance criteria in nuclear
power reactor combined licenses. In any
initial decision under § 52.103(g) or
§ 53.1452(g) of this chapter with respect
to whether acceptance criteria have
been or will be met, the presiding officer
shall make findings of fact and
conclusions of law on the matters put
into controversy by the parties, and any
matter designated by the Commission to
be decided by the presiding officer.
Matters not put into controversy by the
parties but identified by the presiding
officer as matters requiring further
examination, shall be referred to the
Commission for its determination; the
Commission may, in its discretion, treat
any of these referred matters as a request
for action under § 2.206 and process the
matter in accordance with § 52.103(f) or
§ 53.1452(f) of this chapter.
(d) Initial decision—manufacturing
license under 10 CFR parts 52 or 53. (1)
Matters in controversy; presiding officer
consideration of matters not put in
controversy by parties. In any initial
decision in a contested proceeding on
an application for a manufacturing

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Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules
license under subpart C of part 52 or
subpart H of part 53 of this chapter
(including an amendment to or renewal
of a manufacturing license), the
presiding officer shall make findings of
fact and conclusions of law on the
matters put into controversy by the
parties and any matter designated by the
Commission to be decided by the
presiding officer. The presiding officer
also shall make findings of fact and
conclusions of law on any matter not
put into controversy by the parties, but
only to the extent that the presiding
officer determines that a serious safety,
environmental, or common defense and
security matter exists, and the
Commission approves of an
examination of and decision on the
matter upon its referral by the presiding
officer under, inter alia, the provisions
of §§ 2.323 and 2.341.
(2) Presiding officer initial decision
and issuance of permit or license.
(i) In a contested proceeding for the
initial issuance or renewal of a
manufacturing license under subpart C
of part 52 or subpart H of part 53 of this
chapter, or the amendment of a
manufacturing license, the Commission
or the Director, Office of Nuclear
Reactor Regulation, as appropriate, after
making the requisite findings, shall
issue, deny, or appropriately condition
the permit or license in accordance with
the presiding officer’s initial decision
once that decision becomes effective.
(ii) In a contested proceeding for the
initial issuance or renewal of a
manufacturing license under subpart C
of part 52 or subpart H of part 53 of this
chapter, or the amendment of a
manufacturing license, the Commission
or the Director, Office of Nuclear
Reactor Regulation, as appropriate
(appropriate official), may issue the
license, permit, or license amendment
in accordance with § 2.1202(a) or
§ 2.1403(a) before the presiding officer’s
initial decision becomes effective. If,
however, the presiding officer’s initial
decision becomes effective before the
license, permit, or license amendment is
issued under § 2.1202 or § 2.1403, then
the Commission or the Director, Office
of Nuclear Reactor Regulation, as
appropriate, shall issue, deny, or
appropriately condition the license,
permit, or license amendment in
accordance with the presiding officer’s
initial decision.
*
*
*
*
*
(f) Immediate effectiveness of certain
presiding officer decisions. A presiding
officer’s initial decision directing the
issuance or amendment of a limited
work authorization under § 50.10 or
§ 53.1130 of this chapter; an early site

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permit under subpart A of part 52 or
under subpart H of part 53 of this
chapter; a construction permit or
construction authorization under part
50 or part 53 of this chapter; an
operating license under part 50 or part
53 of this chapter; a combined license
under subpart C of part 52 or subpart H
or part 53 of this chapter; a
manufacturing license under subpart F
of part 52 or subpart H of part 53 of this
chapter; a renewed license under part
53 or part 54 of this chapter; or a license
under part 72 of this chapter to store
spent fuel in an independent spent fuel
storage facility (ISFSI) or a monitored
retrievable storage installation (MRS);
an initial decision directing issuance of
a license under part 61 of this chapter;
or an initial decision under § 52.103(g)
or § 53.1452(g) of this chapter that
acceptance criteria in a combined
license have been met, is immediately
effective upon issuance unless the
presiding officer finds that good cause
has been shown by a party why the
initial decision should not become
immediately effective.
*
*
*
*
*
(i) Issuance of authorizations,
permits, and licenses—production and
utilization facilities. The Commission or
the Director, Office of Nuclear Reactor
Regulation, as appropriate, shall issue a
limited work authorization under
§ 50.10 or § 53.1130 of this chapter; an
early site permit under subpart A of part
52 or subpart H of part 53 of this
chapter; a construction permit or
construction authorization under part
50 or part 53 of this chapter; an
operating license under part 50 or part
53 of this chapter; a combined license
under subpart C of part 52 or part 53 of
this chapter; or a manufacturing license
under subpart F of part 52 or part 53 of
this chapter within 10 days from the
date of issuance of the initial decision:
(1) If the Commission or the Director
has made all findings necessary for
issuance of the authorization, permit or
license, not within the scope of the
initial decision of the presiding officer;
and
(2) Notwithstanding the pendency of
a petition for reconsideration under
§ 2.345, a petition for review under
§ 2.341, or a motion for stay under
§ 2.342, or the filing of a petition under
§ 2.206.
(j) Issuance of finding on acceptance
criteria under 10 CFR 52.103 or 10 CFR
53.1452. The Commission or the
Director, Office of Nuclear Reactor
Regulation, as appropriate, shall make
the finding under § 52.103(g) or
§ 53.1452(g) of this chapter, that
acceptance criteria in a combined

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87021

license are met within 10 days from the
date of the presiding officer’s initial
decision:
(1) If the Commission or the Director
is otherwise able to make the finding
under § 52.103(g) or § 53.1452(g) of this
chapter, that the prescribed acceptance
criteria are met for those acceptance
criteria not within the scope of the
initial decision of the presiding officer;
(2) If the presiding officer’s initial
decision—with respect to contentions
that the prescribed acceptance criteria
have not been met—finds that those
acceptance criteria have been met, and
the Commission or the Director
thereafter is able to make the finding
that those acceptance criteria are met;
(3) If the presiding officer’s initial
decision—with respect to contentions
that the prescribed acceptance criteria
will not be met—finds that those
acceptance criteria will be met, and the
Commission or the Director thereafter is
able to make the finding that those
acceptance criteria are met; and
(4) Notwithstanding the pendency of
a petition for reconsideration under
§ 2.345, a petition for review under
§ 2.341, or a motion for stay under
§ 2.342, or the filing of a petition under
§ 2.206.
*
*
*
*
*
§ 2.341

[Amended]

19. In § 2.341(a)(1), remove the phrase
‘‘§ 52.103(c)’’ and add in its place the
phrase ‘‘§ 52.103(c) or § 53.1452(c)’’.

■

§ 2.400

[Amended]

20. In § 2.400, remove the phrase
‘‘parts 50 or 52’’ and add in its place the
phrase ‘‘part 50 or part 52, or
§ 53.1470’’.
■ 21. In § 2.401, revise the section
heading and paragraph (a) to read as
follows:
■

§ 2.401 Notice of hearing on construction
permit or combined license applications
pursuant to appendix N of 10 CFR parts 50,
52, or 53.

(a) In the case of applications under
appendix N of part 50 or § 53.1470 of
this chapter for construction permits for
nuclear power reactors of the type
described in § 50.22 or part 53 of this
chapter, or applications under appendix
N of part 52 or § 53.1470 of this chapter
for combined licenses, the Secretary
will issue notices of hearing pursuant to
§ 2.104.
*
*
*
*
*
■ 22. In § 2.402, revise paragraph (a) to
read as follows:
§ 2.402 Separate hearings on separate
issues; consolidation of proceedings.

(a) In the case of applications under
appendix N of part 50 or § 53.1470 of

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this chapter for construction permits for
nuclear power reactors of a type
described in 10 CFR 50.22 or part 53, or
applications pursuant to appendix N of
part 52 or § 53.1470 of this chapter for
combined licenses, the Commission or
the presiding officer may order separate
hearings on particular phases of the
proceeding, such as matters related to
the acceptability of the design of the
reactor in the context of the site
parameters postulated for the design or
environmental matters.
*
*
*
*
*

§ 2.643 Acceptance and docketing of
application for limited work authorization.

24. In § 2.404, remove the phrase
‘‘appendix N of part 50’’ and add in its
place the phrase ‘‘appendix N to part 50
or § 53.1470’’.

*
*
*
*
(b) The Director will accept for
docketing part one of an application for
a construction permit for a utilization
facility which is subject to § 51.20(b) of
this chapter and is of the type specified
in § 50.21(b)(2) or (3) or § 50.22 or part
53 of this chapter or an application for
a combined license where part one of
the application as described in
§ 2.101(a)(9) is complete. Part one will
not be considered complete unless it
contains the information required by
§ 50.10(d)(3) or § 53.1130(a)(3) of this
chapter. Upon assignment of a docket
number, the procedures in § 2.101(a)(3)
and (4) relating to formal docketing and
the submission and distribution of
additional copies of the application
must be followed.
*
*
*
*
*

§ 2.405

§ 2.645

§ 2.403

[Amended]

23. In § 2.403, remove the phrase
‘‘appendix N of part 50’’ and add in its
place the phrase ‘‘appendix N to part 50
or § 53.1470’’.

■

§ 2.404

[Amended]

■

[Amended]

§ 2.406

[Amended]

26. In § 2.406, remove the phrase
‘‘appendix N of parts 50 or 52’’ and add
in its place the phrase ‘‘appendix N to
part 50 or part 52 or § 53.1470’’.

■

§ 2.500

[Amended]

27. In § 2.500, remove the phrase
‘‘subpart F of part 52’’ and add in its
place the phrase ‘‘subpart F of part 52
or subpart H of part 53’’.
■ 28. In § 2.501, revise the section
heading and paragraph (a) introductory
text to read as follows:
■

§ 2.501 Notice of hearing on application
under 10 CFR parts 52 or 53 for a license
to manufacture nuclear power reactors.

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§ 2.649

(1) An application to construct and/or
operate a production or utilization
facility (including an application for a
limited work authorization under
§§ 50.12 or 53.1130 of this chapter, or an
application for a combined license
under subpart C of 10 CFR part 52, or
under subpart H of 10 CFR part 53;
(2) An application for an early site
permit under subpart A of 10 CFR part
52 or under subpart H of 10 CFR part
53;
(3) An application for a
manufacturing license under subpart F
of 10 CFR part 52 or under subpart H
of 10 CFR part 53;
*
*
*
*
*
(6) Production or utilization facility
licensing actions that involve significant
hazards considerations as defined in
§§ 50.92 or 53.1520 of this chapter.
*
*
*
*
*
§ 2.1301

§ 2.1403

§ 2.800

§ 2.1502

32. In § 2.800, amend paragraphs (c)
and (d) by removing the phrase ‘‘subpart
B of part 52’’ and adding in its place the
phrase ‘‘subpart B of part 52 or subpart
H of part 53’’.

■

[Amended]

33. In § 2.801, remove the phrase
‘‘subpart B of part 52’’ and add in its
place the phrase ‘‘subpart B of part 52
or subpart H of part 53’’.

§ 2.813

[Amended]

34. In § 2.813(a), remove the phrase
‘‘parts 50, 52, and 100’’ and add in its
place the phrase ‘‘parts 50, 52, 53, and
100’’.

■

§ 2.1103

[Amended]

35. In § 2.1103, remove the phrase
‘‘part 50 of this chapter’’ and add in its
place the phrase ‘‘parts 50 or 53 of this
chapter’’.
■ 36. In § 2.1202, revise paragraphs
(a)(1) through (3) and (a)(6) to read as
follows:
■

§ 2.1202

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Authority and role of NRC staff.

(a) * * *

Frm 00106

Fmt 4701

Sfmt 4702

[Amended]

39. In § 2.1500, remove the phrase
‘‘subpart B of part 52’’ and add in its
place the phrase ‘‘subpart B of part 52
or under subpart H of part 53’’.

31. In § 2.649, remove the phrase ‘‘10
CFR 50.10(d)’’ and add in its place the
phrase ‘‘10 CFR 50.10(d) or 10 CFR
53.1130(a)’’.
[Amended]

[Amended]

38. In § 2.1403, remove the phrase ‘‘10
CFR 50.92’’ and add in its place the
phrase ‘‘10 CFR 50.92 or 10 CFR
53.1520’’.

■

■

■

[Amended]

37. In § 2.1301(b), remove ‘‘part 50
and part 52’’ and add in its place ‘‘parts
50, 52, and 53’’.

■

§ 2.1500

[Amended]

■

1 The thirty-day (30) requirement of this
paragraph is not applicable to a notice of the
time and place of hearing published by the
presiding officer after notice of hearing
described in this section has been published.

18:06 Oct 30, 2024

[Amended]

30. In § 2.645, in paragraph (a),
remove the phrase ‘‘§ 50.33(a) through
(f) of this chapter’’ and add in its place
the phrase ‘‘§§ 50.33(a) through (f),
53.1109, and 53.1306(a) or 53.1413 of
this chapter, as applicable,’’.

§ 2.801

(a) In the case of an application under
subpart F of part 52 or subpart H of part
53 of this chapter for a license to
manufacture nuclear power reactors of
the type described in § 50.22 or part 53
of this chapter to be operated at sites not
identified in the license application, the
Secretary will issue a notice of hearing
to be published in the Federal Register
at least 30 days before the date set for
hearing in the notice.1 The notice shall
be issued as soon as practicable after the
application has been docketed. The
notice will state:
*
*
*
*
*

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*

■

25. In § 2.405, remove the phrase
‘‘part 52’’ and add in its place the
phrase ‘‘part 52 or part 53’’.

■

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29. In § 2.643, revise paragraph (b) to
read as follows:

■

[Amended]

40. In § 2.1502, in paragraph (a),
remove the phrase ‘‘§ 52.51(b)’’ and add
in its place the phrase ‘‘§§ 52.51(b) or
53.1242(b)(2)’’; and in paragraph (b)(1),
wherever it appears, remove the phrase
‘‘§ 52.51(a)’’ and add in its place the
phrase ‘‘§§ 52.51(a) or 53.1242(b)’’.

■

PART 10—CRITERIA AND
PROCEDURES FOR DETERMINING
ELIGIBILITY FOR ACCESS TO
RESTRICTED DATA OR NATIONAL
SECURITY INFORMATION OR AN
EMPLOYMENT CLEARANCE
41. The authority citation for part 10
continues to read as follows:

■

Authority: Atomic Energy Act of 1954,
secs. 145, 161 (42 U.S.C. 2165, 2201); Energy
Reorganization Act of 1974, sec. 201 (42
U.S.C. 5841); E.O. 10450, 18 FR 2489, 3 CFR,
1949–1953 Comp., p. 936, as amended; E.O.
10865, 25 FR 1583, 3 CFR, 1959–1963 Comp.,
p. 398, as amended; E.O. 12968, 60 FR 40245,
3 CFR, 1995 Comp., p. 391.
§ 10.1

[Amended]

42. In § 10.1, in paragraph (a)(3)
remove the phrase ‘‘under part 52’’ and

■

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Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules
add in its place the phrase ‘‘under parts
52 or 53’’.
§ 10.2

[Amended]

43. In § 10.2, in paragraph (b),
wherever it appears, remove the phrase
‘‘under part 52’’ and add in its place the
phrase ‘‘under parts 52 or 53’’.

■

PART 11—CRITERIA AND
PROCEDURES FOR DETERMINING
ELIGIBILITY FOR ACCESS TO OR
CONTROL OVER SPECIAL NUCLEAR
MATERIAL

Authority: Atomic Energy Act of 1954,
secs. 161, 223 (42 U.S.C. 2201, 2273); Energy
Reorganization Act of 1974, sec. 201 (42
U.S.C. 5841); 44 U.S.C. 3504 note. Section
11.15(e) also issued under 31 U.S.C. 9701; 42
U.S.C. 2214.
[Amended]

45. In § 11.7, in the introductory text,
remove the phrase ‘‘parts 10, 25, 50, 70,
72, 73, and 95 of this chapter’’ and add
in its place the phrase ‘‘parts 10, 25, 50,
53, 70, 72, 73, and 95 of this chapter’’.

■

PART 19—NOTICES, INSTRUCTIONS
AND REPORTS TO WORKERS:
INSPECTION AND INVESTIGATIONS
46. The authority citation for part 19
continues to read as follows:

■

Authority: Atomic Energy Act of 1954,
secs. 53, 63, 81, 103, 104, 161, 223, 234, 1701
(42 U.S.C. 2073, 2093, 2111, 2133, 2134,
2201, 2273, 2282, 2297f); Energy
Reorganization Act of 1974, secs. 201, 211,
401 (42 U.S.C. 5841, 5851, 5891); 44 U.S.C.
3504 note.

47. In § 19.2, revise paragraph (a) to
read as follows:

■

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§ 19.2

Scope.

(a) * * *
(1) All persons who receive, possess,
use, or transfer material licensed by the
NRC under the regulations in parts 30
through 36, 39, 40, 60, 61, 63, 70, or 72
of this chapter, including persons
licensed to operate a production or
utilization facility under part 50, part
52, or part 53 of this chapter, persons
licensed to possess power reactor spent
fuel in an independent spent fuel
storage installation (ISFSI) under part 72
of this chapter, and in accordance with
10 CFR 76.60 to persons required to
obtain a certificate of compliance or an
approved compliance plan under part
76 of this chapter;
(2) All applicants for and holders of
licenses (including construction permits
and early site permits) under parts 50,
52, 53, and 54 of this chapter;
(3) All applicants for and holders of
a standard design approval under

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§ 19.3

*
*
*
*
License means a license issued under
the regulations in parts 30 through 36,
39, 40, 60, 61, 63, 70, or 72 of this
chapter, including licenses to
manufacture, construct and/or operate a
production or utilization facility under
parts 50, 52, 53, or 54 of this chapter.
*
*
*
*
*
Regulated entities means any
individual, person, organization, or
corporation that is subject to the
regulatory jurisdiction of the NRC,
including (but not limited to) an
applicant for or holder of a standard
design approval under subpart E of part
52 or under subpart H of part 53 of this
chapter or a standard design
certification under subpart B of part 52
or under subpart H of part 53 of this
chapter.
*
*
*
*
*
§ 19.11

[Amended]

49. In § 19.11, in paragraph (a)
introductory text, paragraph (b)
introductory text, and paragraph (e)(1),
remove the phrase ‘‘of part 52’’
wherever may appears and add in its
place the phrase ‘‘of part 52 or under
subpart H of part 53’’.

■

§ 19.14

[Amended]

50. In § 19.14, in paragraph (a),
wherever it may appear, remove the
phrase ‘‘of part 52’’ and add in its place
the phrase ‘‘of part 52 or under subpart
H of part 53’’.

■

§ 19.20

[Amended]

51. In § 19.20, add the number ‘‘53,’’
in sequential order.

■

PART 20—STANDARDS FOR
PROTECTION AGAINST RADIATION
52. The authority citation for part 20
continues to read as follows:

■

Authority: Atomic Energy Act of 1954,
secs. 11, 53, 63, 65, 81, 103, 104, 161, 170H,
182, 186, 223, 234, 274, 1701 (42 U.S.C. 2014,
2073, 2093, 2095, 2111, 2133, 2134, 2201,
2210h, 2232, 2236, 2273, 2282, 2021, 2297f);
Energy Reorganization Act of 1974, secs. 201,
202 (42 U.S.C. 5841, 5842); Low-Level
Radioactive Waste Policy Amendments Act

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of 1985, sec. 2 (42 U.S.C. 2021b); 44 U.S.C.
3504 note.
§ 20.1002

Sfmt 4702

[Amended]

53. In § 20.1002, remove the phrase
‘‘parts 30 through 36, 39, 40, 50, 52, 60,
61, 63, 70, or 72 of this chapter’’ and
add in its place the phrase ‘‘parts 30
through 36, 39, 40, 50, 52, 53, 60, 61,
63, 70, or 72 of this chapter’’.
■ 54. In § 20.1003, revise the definition
for ‘‘License’’ to read as follows:
■

§ 20.1003

Definitions.

*

44. The authority citation for part 11
continues to read as follows:

■

§ 11.7

subpart E of part 52 or under subpart H
of part 53 of this chapter; and
(4) All applicants for a standard
design certification under subpart B of
part 52 or under subpart H of part 53 of
this chapter, and those (former)
applicants whose designs have been
certified under that subpart.
*
*
*
*
*
■ 48. In § 19.3, revise the definitions for
‘‘License’’ and ‘‘Regulated entities’’ to
read as follows:

87023

Definitions.

*

*
*
*
*
License means a license issued under
the regulations in parts 30 through 36,
39, 40, 50, 53, 60, 61, 63, 70, or 72 of
this chapter.
*
*
*
*
*
§ 20.1101

[Amended]

55. In § 20.1101, in paragraph (d),
remove the phrase ‘‘subject to § 50.34a’’
and add in its place the phrase ‘‘subject
to §§ 50.34a or 53.260 of this chapter’’.

■

§ 20.1401

[Amended]

56. Amend § 20.1401 by:
a. In paragraph (a), removing the
phrase ‘‘parts 30, 40, 50, 52, 60, 61, 63,
70, and 72 of this chapter’’, and adding
in its place the phrase ‘‘parts 30, 40, 50,
52, 53, 60, 61, 63, 70, and 72 of this
chapter’’; and
■ b. In paragraphs (a) and (c) removing
the phrase ‘‘in accordance with § 50.83’’
and adding in its place the phrase ‘‘in
accordance with §§ 50.83 or 53.1080’’.
■ 57. In § 20.1403, revise paragraph (d)
introductory text to read as follows:
■
■

§ 20.1403 Criteria for license termination
under restricted conditions.

*

*
*
*
*
(d) The licensee has submitted a
decommissioning plan or License
Termination Plan (LTP) to the
Commission indicating the licensee’s
intent to decommission in accordance
with §§ 30.36(d), 40.42(d), 50.82 (a) and
(b), subpart G of part 53, 70.38(d), or
72.54 of this chapter, and specifying
that the licensee intends to
decommission by restricting use of the
site. The licensee shall document in the
LTP or decommissioning plan how the
advice of individuals and institutions in
the community who may be affected by
the decommissioning has been sought
and incorporated, as appropriate,
following analysis of that advice.
*
*
*
*
*
■ 58. In § 20.1404, revise paragraph
(a)(4) introductory text to read as
follows:
§ 20.1404 Alternate criteria for license
termination.

(a) * * *

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(4) Has submitted a decommissioning
plan or License Termination Plan (LTP)
to the Commission indicating the
licensee’s intent to decommission in
accordance with § 30.36(d), 40.42(d),
50.82 (a) and (b), subpart G of part 53,
70.38(d), or 72.54 of this chapter, and
specifying that the licensee proposes to
decommission by use of alternate
criteria. The licensee shall document in
the decommissioning plan or LTP how
the advice of individuals and
institutions in the community who may
be affected by the decommissioning has
been sought and addressed, as
appropriate, following analysis of that
advice. In seeking such advice, the
licensee shall provide for:
*
*
*
*
*
§ 20.1406

[Amended]

59. In § 20.1406, in paragraphs (a) and
(b), wherever it appears, remove the
phrase ‘‘under part 52’’ and add in its
place the phrase ‘‘under parts 52 or 53’’.
■ 60. In § 20.1501, revise paragraph (b)
to read as follows:
■

§ 20.1501

General.

*

*
*
*
*
(b) Notwithstanding § 20.2103(a) of
this part, records from surveys
describing the location and amount of
subsurface residual radioactivity
identified at the site must be kept with
records important for decommissioning,
and such records must be retained in
accordance with § 30.35(g), § 40.36(f),
§ 50.75(g), subpart G of part 53,
§ 70.25(g), or § 72.30(d) of this chapter,
as applicable.
*
*
*
*
*
§ 20.1905

[Amended]

61. In § 20.1905, in paragraph (g)
introductory text, remove the phrase
‘‘Parts 50 or 52’’ and add in its place the
phrase ‘‘parts 50, 52, or 53’’.
■ 62. In § 20.2004, revise paragraph
(b)(1) to read as follows:
■

§ 20.2004 Treatment or disposal by
incineration.

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*

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(g) or 53.1640(b), (c), (d), and (e) of this
chapter, and must include the
information required by paragraph (b) of
this section. Occurrences reported
under §§ 50.73 or 53.1640 of this
chapter need not be reported by a
duplicate report under paragraph (a) of
this section.
*
*
*
*
*
§ 20.2206

[Amended]

66. In § 20.2206, in paragraph (a)(1),
remove the phrase ‘‘or § 50.22’’ and add
in its place the phrase ‘‘, § 50.22, or part
53’’.

■

PART 21—REPORTING OF DEFECTS
AND NONCOMPLIANCE
67. The authority citation for part 21
continues to read as follows:

§ 20.2201 Reports of theft or loss of
licensed material.

■

(a) * * *
(2) * * *
(i) Licensees having an installed
Emergency Notification System shall
make the reports to the NRC Operations
Center under §§ 50.72 or 53.1630 of this
chapter, and
*
*
*
*
*
(b) * * *
(2) * * *
(i) For holders of an operating license
for a nuclear power plant, the events
included in paragraph (b) of this section
must be reported under the procedures
described in §§ 50.73(b), (c), (d), (e), and
(g) or 53.1640(b), (c), (d), and (e) of this
chapter and must include the
information required in paragraph (b)(1)
of this section, and
*
*
*
*
*
(c) A duplicate report is not required
under paragraph (b) of this section if the
licensee is also required to submit a
report pursuant to §§ 30.55(c), 37.57,
37.81, 40.64(c), 50.72, 50.73, 53.1630,
53.1640, 70.52, 73.27(b), 73.67(e)(3)(vii),
73.67(g)(3)(iii), 73.1205, or 150.19(c) of
this chapter.
*
*
*
*
*

Authority: Atomic Energy Act of 1954,
secs. 53, 63, 81, 103, 104, 161, 223, 234, 1701
(42 U.S.C. 2073, 2093, 2111, 2133, 2134,
2201, 2273, 2282, 2297f); Energy
Reorganization Act of 1974, secs. 201, 206
(42 U.S.C. 5841, 5846); Nuclear Waste Policy
Act of 1982, secs. 135, 141 (42 U.S.C. 10155,
10161); 44 U.S.C. 3504 note.

§ 20.2202

*
*
*
*
(b)(1) Waste oils (petroleum derived
or synthetic oils used principally as
lubricants, coolants, hydraulic or
insulating fluids, or metalworking oils)
that have been radioactively
contaminated in the course of the
operation or maintenance of a nuclear
power reactor licensed under parts 50 or
53 of this chapter may be incinerated on
the site where generated provided that
the total radioactive effluents from the
facility, including the effluents from
such incineration, conform to the
requirements of appendix I to part 50 or
§ 53.425(d) of this chapter and the

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effluent release limits contained in
applicable license conditions other than
effluent limits specifically related to
incineration of waste oil. The licensee
shall report any changes or additions to
the information supplied under
§§ 50.34, 50.34a, or under subpart H of
part 53 of this chapter associated with
this incineration pursuant to §§ 50.71 or
53.1620 of this chapter, as appropriate.
The licensee shall also follow the
procedures of §§ 50.59 or 53.1565 of this
chapter with respect to such changes to
the facility or procedures.
*
*
*
*
*
■ 63. In § 20.2201, revise paragraphs
(a)(2)(i), (b)(2)(i), and (c) to read as
follows:

[Amended]

64. In § 20.2202, in paragraph (d)(1),
remove the phrase ‘‘10 CFR 50.72’’ and
add in its place the phrase ‘‘§§ 50.72 or
53.1630 of this chapter;’’.
■ 65. In § 20.2203, revise paragraph (c)
to read as follows:
■

§ 20.2203 Reports of exposures, radiation
levels, and concentrations of radioactive
material exceeding the constraints or limits.

*

*
*
*
*
(c) For holders of an operating license
or a combined license for a nuclear
power plant, the occurrences included
in paragraph (a) of this section must be
reported under the procedures
described in §§ 50.73(b), (c), (d), (e), and

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68. In § 21.2, revise paragraphs (a)(2)
through (4), (b), and (c) to read as
follows:

■

§ 21.2

Scope.

(a) * * *
(1) * * *
(2) Each individual, corporation,
partnership, or other entity doing
business within the United States, and
each director and responsible officer of
such an organization, that constructs a
production or utilization facility
licensed for manufacture, construction,
or operation under parts 50, 52, or 53 of
this chapter, an ISFSI for the storage of
spent fuel licensed under part 72 of this
chapter, an MRS for the storage of spent
fuel or high-level radioactive waste
under part 72 of this chapter, or a
geologic repository for the disposal of
high-level radioactive waste under parts
60 or 63 of this chapter; or supplies
basic components for a facility or
activity licensed, other than for export,
under parts 30, 40, 50, 52, 53, 60, 61, 63,
70, 71, or 72 of this chapter;
(3) Each individual, corporation,
partnership, or other entity doing
business within the United States, and
each director and responsible officer of
such an organization, applying for a
design certification rule under parts 52
or 53 of this chapter; or supplying basic
components with respect to that design
certification, and each individual,
corporation, partnership, or other entity
doing business within the United States,
and each director and responsible
officer of such an organization, whose

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application for design certification has
been granted under parts 52 or 53 of this
chapter, or who has supplied or is
supplying basic components with
respect to that design certification;
(4) Each individual, corporation,
partnership, or other entity doing
business within the United States, and
each director and responsible officer of
such an organization, applying for or
holding a standard design approval
under parts 52 or 53 of this chapter; or
supplying basic components with
respect to a standard design approval
under parts 52 or 53 of this chapter;
(b) For persons licensed to construct
a facility under either a construction
permit issued under §§ 50.23 or 53.1333
of this chapter or a combined license
under parts 52 or 53 of this chapter (for
the period of construction until the date
that the Commission makes the finding
under §§ 52.103(g) or 53.1452(g) of this
chapter), or to manufacture a facility
under parts 52 or 53 of this chapter,
evaluation of potential defects and
failures to comply and reporting of
defects and failures to comply under
§§ 50.55(e) or 53.605 of this chapter
satisfies each person’s evaluation,
notification, and reporting obligation to
report defects and failures to comply
under this part and the responsibility of
individual directors and responsible
officers of these licensees to report
defects under Section 206 of the Energy
Reorganization Act of 1974.
(c) For persons licensed to operate a
nuclear power plant under part 50, part
52, or part 53 of this chapter, evaluation
of potential defects and appropriate
reporting of defects under §§ 50.72,
50.73, 53.1630, 53.1640, or 73.1200 and
73.1205 of this chapter, satisfies each
person’s evaluation, notification, and
reporting obligation to report defects
under this part, and the responsibility of
individual directors and responsible
officers of these licensees to report
defects under Section 206 of the Energy
Reorganization Act of 1974.
*
*
*
*
*
■ 69. In § 21.3, revise the definitions for
‘‘Basic component’’, ‘‘Commercial grade
item’’, ‘‘Critical characteristics’’,
‘‘Dedicating entity’’, ‘‘Dedication’’,
‘‘Defect’’, and ‘‘Substantial safety
hazard’’ to read as follows:
§ 21.3

Definitions.

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*

*
*
*
*
Basic component. (1)(i) When applied
to nuclear power plants licensed under
part 53 of this chapter, basic component
means a safety-related structure, system,
or component (SSC), or part thereof, and
when applied to nuclear power plants
licensed under parts 50 or 52, of this
chapter, basic component means an

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SSC, or part thereof that affects its safety
function necessary to assure:
(A) The integrity of the reactor coolant
pressure boundary;
(B) The capability to shut down the
reactor and maintain it in a safeshutdown condition; or
(C) The capability to prevent or
mitigate the consequences of accidents
which could result in potential offsite
exposures comparable to those referred
to in §§ 50.34(a)(1), 50.67(b)(2), or
100.11 of this chapter, as applicable.
(ii) Basic components are items
designed and manufactured under a
quality assurance program complying
with appendix B to part 50 of this
chapter, or commercial grade items
which have successfully completed the
dedication process.
(2) When applied to standard design
certifications and approvals under part
53 of this chapter, basic component
means the design or procurement
information approved or to be approved
within the scope of the design
certification or approval for a safetyrelated SSC, or part thereof. When
applied to standard design certifications
under subpart B of part 52 of this
chapter and standard design approvals
under part 52 of this chapter, basic
component means the design or
procurement information approved or to
be approved within the scope of the
design certification or approval for an
SSC, or part thereof, that affects its
safety function necessary to assure:
(i) The integrity of the reactor coolant
pressure boundary;
(ii) The capability to shut down the
reactor and maintain it in a safeshutdown condition; or
(iii) The capability to prevent or
mitigate the consequences of accidents
which could result in potential offsite
exposures comparable to those referred
to in §§ 50.34(a)(1), 50.67(b)(2), or
100.11 of this chapter, as applicable.
(3) When applied to other facilities
and other activities licensed under 10
CFR parts 30, 40, 50 (other than nuclear
power plants), 60, 61, 63, 70, 71, or 72
of this chapter, basic component means
a structure, system, or component, or
part thereof, that affects their safety
function, that is directly procured by the
licensee of a facility or activity subject
to the regulations in this part and in
which a defect or failure to comply with
any applicable regulation in this
chapter, order, or license issued by the
Commission could create a substantial
safety hazard.
(4) In all cases, basic component
includes safety-related design, analysis,
inspection, testing, fabrication,
replacement of parts, or consulting
services that are associated with the

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87025

component hardware, design
certification, design approval, or
information in support of an early site
permit application under part 52 or part
53 of this chapter, whether these
services are performed by the
component supplier or others.
Commercial grade item. (1) When
applied to nuclear power plants
licensed under parts 50 or 53 of this
chapter, commercial grade item means
an SSC, or part thereof that affects its
safety function, that was not designed
and manufactured as a basic
component. Commercial grade items do
not include items where the design and
manufacturing process require inprocess inspections and verifications to
ensure that defects or failures to comply
are identified and corrected (i.e., one or
more critical characteristics of the item
cannot be verified).
(2) When applied to facilities and
activities licensed pursuant to parts 30,
40, 50 (other than nuclear power
plants), 60, 61, 63, 70, 71, or 72 of this
chapter, commercial grade item means
an item that is:
(i) Not subject to design or
specification requirements that are
unique to those facilities or activities;
(ii) Used in applications other than
those facilities or activities; and
(iii) To be ordered from the
manufacturer/supplier on the basis of
specifications set forth in the
manufacturer’s published product
description (for example, a catalog).
*
*
*
*
*
Critical characteristics. When applied
to nuclear power plants licensed under
parts 50, 52, or 53 of this chapter,
critical characteristics are those
important design, material, and
performance characteristics of a
commercial grade item that, once
verified, will provide reasonable
assurance that the item will perform its
intended safety function.
Dedicating entity. When applied to
nuclear power plants licensed under
parts 50, 52, or 53 of this chapter,
dedicating entity means the
organization that performs the
dedication process. Dedication may be
performed by the manufacturer of the
item, a third-party dedicating entity, or
the licensee itself. The dedicating entity,
under § 21.21(c) of this part, is
responsible for identifying and
evaluating deviations, reporting defects
and failures to comply for the dedicated
item, and maintaining auditable records
of the dedication process.
Dedication. (1) When applied to
nuclear power plants licensed pursuant
to 10 CFR parts 30, 40, 50, 53, or 60,
dedication is an acceptance process

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undertaken to provide reasonable
assurance that a commercial grade item
to be used as a basic component will
perform its intended safety function
and, in this respect, is deemed
equivalent to an item designed and
manufactured under a 10 CFR part 50,
appendix B, quality assurance program.
This assurance is achieved by
identifying the critical characteristics of
the item and verifying their
acceptability by inspections, tests, or
analyses performed by the purchaser or
third-party dedicating entity after
delivery, supplemented as necessary by
one or more of the following:
commercial grade surveys; product
inspections or witness at holdpoints at
the manufacturer’s facility, and analysis
of historical records for acceptable
performance. In all cases, the dedication
process must be conducted under the
applicable provisions of 10 CFR part 50,
appendix B. The process is considered
complete when the item is designated
for use as a basic component.
(2) When applied to facilities and
activities licensed pursuant to 10 CFR
parts 30, 40, 50 (other than nuclear
power plants), 60, 61, 63, 70, 71, or 72,
dedication occurs after receipt when
that item is designated for use as a basic
component.
Defect means:
(1) A deviation in a basic component
delivered to a purchaser for use in a
facility or an activity subject to the
regulations in this part if, on the basis
of an evaluation, the deviation could
create a substantial safety hazard;
(2) The installation, use, or operation
of a basic component containing a
defect as defined in this section;
(3) A deviation in a portion of a
facility subject to the early site permit,
standard design certification, standard
design approval, construction permit,
combined license or manufacturing
licensing requirements of parts 50, 52,
or 53 of this chapter, provided the
deviation could, on the basis of an
evaluation, create a substantial safety
hazard and the portion of the facility
containing the deviation has been
offered to the purchaser for acceptance;
(4) A condition or circumstance
involving a basic component that could
contribute to the exceeding of a safety
limit, as defined in the technical
specifications of a license for operation
issued under part 50, part 52, or part 53
of this chapter; or
(5) An error, omission or other
circumstance in a design certification,
or standard design approval that, on the
basis of an evaluation, could create a
substantial safety hazard.
*
*
*
*
*

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Substantial safety hazard means a
loss of safety function to the extent that
there is a major reduction in the degree
of protection provided to public health
and safety for any facility or activity
licensed or otherwise approved or
regulated by the NRC, other than for
export, under part 30, 40, 50, 52, 53, 60,
61, 63, 70, 71, or 72 of this chapter.
*
*
*
*
*

§ 25.35

§ 21.21

Authority: Atomic Energy Act of 1954,
secs. 53, 103, 104, 107, 161, 223, 234, 1701
(42 U.S.C. 2073, 2133, 2134, 2137, 2201,
2273, 2282, 2297f); Energy Reorganization
Act of 1974, secs. 201, 202 (42 U.S.C. 5841,
5842); 44 U.S.C. 3504 note.

[Amended]

70. Amend § 21.21 by:
■ a. In paragraph (a)(3), removing the
phrase ‘‘under part 52’’ and add in its
place the phrase ‘‘under parts 52 or 53’’;
and
■ b. In paragraphs (d)(1)(i) and (ii)
removing the phrase ‘‘parts 30, 40, 50,
52, 60, 61, 63, 70, 71, or 72 of this
chapter’’ and adding in its place the
phrase ‘‘parts 30, 40, 50, 52, 53, 60, 61,
63, 70, 71, or 72 of this chapter’’.
■

§ 21.51

[Amended]

71. In § 21.51, in paragraphs (a)(4) and
(5) remove the phrase ‘‘of part 52’’ and
add in its place the phrase ‘‘of part 52
or under subpart H of part 53’’.

■

§ 21.61

[Amended]

72. In § 21.61, in paragraph (b) remove
the phrase ‘‘under part 52’’ and add in
its place the phrase ‘‘under parts 52 or
53’’.

■

PART 25—ACCESS AUTHORIZATION
73. The authority citation for part 25
continues to read as follows:

■

Authority: Atomic Energy Act of 1954,
secs. 145, 161, 223, 234 (42 U.S.C. 2165,
2201, 2273, 2282); Energy Reorganization Act
of 1974, sec. 201 (42 U.S.C. 5841); 44 U.S.C.
3504 note; E.O. 10865, 25 FR 1583, as
amended, 3 CFR, 1959–1963 Comp., p. 398;
E.O. 12829, 58 FR 3479, 3 CFR, 1993 Comp.,
p. 570; E.O. 13526, 75 FR 707, 3 CFR, 2009
Comp., p. 298; E.O. 12968, 60 FR 40245, 3
CFR, 1995 Comp., p. 391. Section 25.17(f)
and Appendix A also issued under 31 U.S.C.
9701; 42 U.S.C. 2214.

74. In § 25.5, revise the definition for
‘‘License’’ to read as follows:

■

§ 25.5

*
*
*
*
License means a license issued
pursuant to 10 CFR parts 50, 52, 53, 60,
63, 70, or 72.
*
*
*
*
*
[Amended]

75. In § 25.17, in paragraph (a),
remove the phrase ‘‘under 10 CFR parts
50, 52, 54, 60, 63, 70, 72, or 76’’ and add
in its place the phrase ‘‘under 10 CFR
parts 50, 52, 53, 54, 60, 63, 70, 72, or
76’’.

■

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PART 26—FITNESS FOR DUTY
PROGRAMS
77. The authority citation for part 26
continues to read as follows:

■

78. In § 26.3, revise paragraph (d) and
add paragraph (f) to read as follows:

■

§ 26.3

Scope.

*

*
*
*
*
(d) Contractor/vendors (C/Vs) who
implement FFD programs or program
elements, to the extent that the licensees
and other entities specified in
paragraphs (a) through (c) and (f) of this
section rely on those C/V FFD programs
or program elements to meet the
requirements of this part, shall comply
with the requirements of this part.
*
*
*
*
*
(f) No later than the start of
construction activities, licensees and
other entities that have applied for or
have been issued a license under part 53
of this chapter, other than a
manufacturing license (ML), must
implement the requirements in subpart
M of this part or all the requirements of
this part except subparts K and M.
Holders of an ML under part 53 of this
chapter must implement the
requirements in subpart M or all the
requirements of this part except
subparts K and M, before commencing
activities that assemble a manufactured
reactor.
■ 79. In § 26.4, revise paragraphs (a)
introductory text, (a)(1), (a)(4), (b), (c),
(e) introductory text, (e)(4), (f), (g)
introductory text, and (h) to read as
follows:
§ 26.4 FFD program applicability to
categories of individuals.

Definitions.

*

§ 25.17

[Amended]

76. In § 25.35, in paragraph (a),
wherever it appears, remove the phrase
‘‘under part 52’’ and add in its place the
phrase ‘‘under parts 52 or 53’’.

■

Sfmt 4702

(a) All persons who are granted
unescorted access to nuclear power
reactor protected areas by the licensees
in § 26.3(a) and, as applicable, (c) and
perform the following duties shall be
subject to an FFD program that meets all
of the requirements of this part, except
subpart K of this part, and those persons
who are granted unescorted access to
either nuclear power reactor protected
areas or remote facilities where safetysignificant systems or components may
be operated within the design basis of

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a licensed commercial nuclear plant, by
the licensees and other entities in
§ 26.3(f) and perform the following
duties must be subject to an FFD
program that satisfies the requirements
in subpart M of this part, unless the
licensee or other entity subjects these
individuals to an FFD program that
satisfies all of the requirements of this
part except for those requirements in
subparts K and M:
(1) For persons who are granted
unescorted access by the licensees in
§ 26.3(a) and, as applicable, (c),
operating or onsite directing of the
operation of systems and components
that a risk-informed evaluation process
has shown to be significant to public
health and safety; for those persons who
are granted unescorted access by the
licensees and other entities in § 26.3(f),
operating or directing of the operation of
systems and components that a riskinformed evaluation process has shown
to be significant to public health and
safety;
*
*
*
*
*
(4) For persons who are granted
unescorted access to nuclear power
reactor protected areas by the licensees
in § 26.3(a) and, as applicable, (c),
performing maintenance or onsite
directing of the maintenance of SSCs
that a risk-informed evaluation process
has shown to be significant to public
health and safety; for those persons who
are granted unescorted access to nuclear
power reactor protected areas by the
licensees and other entities in § 26.3(f),
performing maintenance or directing of
the maintenance of SSCs that a riskinformed evaluation process has shown
to be significant to public health and
safety; and
*
*
*
*
*
(b) All persons who are granted
unescorted access to nuclear power
reactor protected areas by the licensees
in § 26.3(a) and, as applicable, (c) and
who do not perform the duties
described in paragraph (a) of this
section shall be subject to an FFD
program that meets all of the
requirements of this part, except
§§ 26.205 through 26.209 and subpart K
of this part. All persons who are granted
unescorted access to a facility licensed
under part 53 of this chapter, and who
do not perform or direct the
performance of the duties described in
§ 26.4(a), must be subject to the
requirements in subpart M of this part,
unless the licensee or other entity
implements an FFD program that
satisfies all of the requirements of this
part, except §§ 26.205 through 26.209
and subparts K and M.

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(c) All persons who are required by a
licensee in § 26.3(a) and, as applicable,
(c) to physically report to the licensee’s
Technical Support Center or Emergency
Operations Facility by licensee
emergency plans and procedures shall
be subject to an FFD program that meets
all of the requirements of this part,
except §§ 26.205 through 26.209 and
subpart K of this part. Also, for licensees
or other entities in § 26.3(f), all persons
without unescorted access to the facility
who make decisions and/or direct
actions regarding plant safety and
security, and all persons who
participate remotely in emergency
response activities or physically report
to the Technical Support Center or
Emergency Operations Facility (or an
equivalent facility), must be subject to
an FFD program that satisfies all of the
requirements described in subpart M of
this part, unless the licensee or other
entity implements an FFD program that
satisfies all of the requirements of this
part, except §§ 26.205 through 26.209
and subparts K and M.
*
*
*
*
*
(e) When construction activities, as
defined in § 26.5, begin, any individual
whose duties for the licensees and other
entities in § 26.3(c) require him or her
to have the following types of access or
perform the following activities at the
location where the nuclear power plant
will be constructed and operated shall
be subject to an FFD program that meets
all of the requirements of this part,
except subparts I, K, and M of this part,
and for any individual whose duties for
the licensees and other entities in
§ 26.3(f) require him or her to have the
following types of access, perform
construction activities as defined in
§ 26.5, or perform the following
activities must be subject to an FFD
program as described in subpart M or an
FFD program that satisfies all of the
requirements of this part, except
subparts I, K, and M:
*
*
*
*
*
(4) Witnesses or determines
inspections, tests, and analyses
certification required under part 52 or
part 53 of this chapter;
*
*
*
*
*
(f) Any individual who is constructing
or directing the construction of safetyor security-related SSCs shall be subject
to an FFD program that meets the
requirements of subpart K, or, if
applicable, subpart M of this part,
unless the licensee or other entity
subjects these individuals to an FFD
program that meets all of the
requirements of this part, except for
subparts I, K, and M of this part.

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(g) All FFD program personnel who
are involved in the day-to-day
operations of the program, as defined by
the procedures of the licensees and
other entities in § 26.3(a) through (c),
and, as applicable, (d) and whose duties
require them to have the following types
of access or perform the following
activities shall be subject to an FFD
program that meets all of the
requirements of this part, except
subparts I, K, and M of this part, and,
at the licensee’s or other entity’s
discretion, subpart C of this part. All
personnel whose duties require them to
have the following types of access or
perform the following activities at
facilities licensed under part 53 of this
chapter must be subject to the
requirements in subpart M or an FFD
program that satisfies all of the
requirements of this part, except
subparts I, K, and M, and, at the
licensee’s or other entity’s discretion,
subpart C of this part:
*
*
*
*
*
(h) Individuals who have applied for
authorization to have the types of access
or perform the activities described in
paragraphs (a) through (d) of this section
shall be subject to §§ 26.31(c)(1),
26.35(b), 26.37, 26.39, and the
applicable requirements of subparts C, E
through H, and M of this part.
*
*
*
*
*
■ 80. Amend § 26.5 by:
■ a. Adding the definitions for
‘‘Biological marker’’ and ‘‘Change’’;
■ b. Revising the definitions for
‘‘Constructing or construction
activities’’ ‘‘Contractor/vendor (C/V)’’;
■ c. Adding the definition of ‘‘Illicit
substance’’;
■ d. Revising the definitions of ‘‘Other
entity’’ and ‘‘Questionable validity’’;
■ e. Adding the definitions of
‘‘Reduction in FFD program
effectiveness’’;
■ f. Revising the definitions of
‘‘Reviewing official’’, ‘‘Safety-related
structures, systems, and components
(SSCs)’’, and ‘‘Security-related SSCs’’;
■ g. Adding the definitions of ‘‘Special
nuclear material’’; and
■ h. Revising the definition of ‘‘Unit
outage’’.
The additions and revisions read as
follows:
§ 26.5

Definitions.

*

*
*
*
*
Biological marker means, for a part 53
licensee implementing subpart M of this
part, an endogenous substance that is
used to validate that the biological
specimen collected for testing was
produced by the donor.
*
*
*
*
*

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Change as used in § 26.603(e) means
an action that results in a modification
of, addition to, or removal from the
licensee’s or other entity’s FFD program.
*
*
*
*
*
Constructing or construction activities
means, for the purposes of this part, the
tasks involved in building a nuclear
power plant that are performed at the
location where the nuclear power plant
will be constructed and operated. These
tasks include fabricating, erecting,
integrating, and testing safety- and
security-related SSCs, and the
installation of their foundations,
including the placement of concrete. For
a licensee or other entity described in
§ 26.3(f), construction is defined in
§ 53.020 of this chapter.
Contractor/vendor (C/V) means any
company, or any individual not
employed by a licensee or other entity
specified in § 26.3(a) through (c) and (f),
who is providing work or services to a
licensee or other entity covered in
§ 26.3(a) through (c) and (f), either by
contract, purchase order, oral
agreement, or other arrangement.
*
*
*
*
*
Illicit substance means a substance
that causes impairment and possible
addiction but is not an illegal drug as
defined in § 26.5.
*
*
*
*
*
Other entity means any corporation,
firm, partnership, limited liability
company, association, C/V, or other
organization who is subject to this part
under § 26.3(a) through (c) and (f) but is
not licensed by the NRC.
*
*
*
*
*
Questionable validity means the
results of validity screening or initial
validity tests at a licensee testing facility
indicating that a urine specimen may be
adulterated, substituted, dilute, or
invalid. For a part 53 licensee or other
entity, questionable validity means the
results of validity screening or initial
validity tests indicating that a biological
specimen obtained from an individual
pursuant to subpart M of this part may
be adulterated, substituted, dilute, or
invalid.
Reduction in FFD program
effectiveness means, for a part 53
licensee or other entity implementing
subpart M of this part, a change or series
of changes to an element of the FFD
program that reduces or eliminates the
licensee’s ability to satisfy or maintain
site-specific FFD program performance
when compared to historical sitespecific performance, the licensee’s
fleet-level program performance, or
industry performance.
*
*
*
*
*

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Reviewing official means an employee
of a licensee or other entity specified in
§ 26.3(a) through (c) and (f), who is
designated by the licensee or other
entity to be responsible for reviewing
and evaluating any potentially
disqualifying FFD information about an
individual, including, but not limited
to, the results of a determination of
fitness, as defined in § 26.189, in order
to determine whether the individual
may be granted or maintain
authorization.
Safety-related structures, systems, and
components (SSCs) means, for part 50 or
part 52 licensees and other entities
described in § 26.3(a) through (d), those
SSCs that are relied on to remain
functional during and following design
basis events to ensure the integrity of
the reactor coolant pressure boundary,
the capability to shut down the reactor
and maintain it in a safe shutdown
condition, or the capability to prevent or
mitigate the consequences of accidents
that could result in potential offsite
exposure comparable to the guidelines
in § 50.34(a)(1) of this chapter. For part
53 licensees and other entities described
in § 26.3(d) and (f), safety-related has
the same meaning as that in § 53.020 of
this chapter.
Security-related SSCs means, for the
purposes of this part, those structures,
systems, and components that the
licensee will rely on to implement the
licensee’s physical security and
safeguards contingency plans that either
are required under part 73 of this
chapter if the licensee is a construction
permit applicant or holder or an early
site permit holder, as described in
§ 26.3(c)(3) through (c)(5), respectively,
or are included in the licensee’s
application if the licensee is a combined
license applicant or holder, as described
in § 26.3(c)(1) and (c)(2), respectively, or
a licensee or other entity described in
§ 26.3(d) or (f).
*
*
*
*
*
Special nuclear material (SNM) has
the same meaning as that in § 70.4 of
this chapter.
*
*
*
*
*
Unit outage means, for the purposes
of this part, for electricity-generation
units, that the reactor unit is
disconnected from the electrical grid.
Unit outage means, for the purposes of
this part, for non-electricity-generation
units, that the reactor unit is
disconnected from the loads to which
its output is supplied under normal
operating conditions.
*
*
*
*
*
■ 81. In § 26.8, revise paragraph (b) to
read as follows:

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§ 26.8 Information collection
requirements: OMB approval.

*

*
*
*
*
(b) The approved information
collection requirements contained in
this part appear in §§ 26.9, 26.27, 26.29,
26.31, 26.33, 26.35, 26.37, 26.39, 26.41,
26.53, 26.55, 26.57, 26.59, 26.61, 26.63,
26.65, 26.67, 26.69, 26.75, 26.77, 26.85,
26.87, 26.89, 26.91, 26.93, 26.95, 26.97,
26.99, 26.101, 26.103, 26.107, 26.109,
26.111, 26.113, 26.115, 26.117, 26.119,
26.125, 26.127, 26.129, 26.135, 26.137,
26.139, 26.153, 26.157, 26.159, 26.163,
26.165, 26.167, 26.168, 26.169, 26.183,
26.185, 26.187, 26.189, 26.202, 26.203,
26.205, 26.207, 26.211, 26.401, 26.403,
26.405, 26.406, 26.407, 26.411, 26.413,
26.415, 26.417, 26.603, 26.604, 26.605,
26.606, 26.607, 26.608, 26.609, 26.611,
26.613, 26.617, 26.619, 26.711, 26.713,
26.715, 26.717, 26.719, and 26.821.
■ 82. Revise § 26.21 to read as follows:
§ 26.21

Fitness-for-duty program.

The licensees and other entities
specified in § 26.3(a) through (c) and (f)
(for those licensees and other entities
that do not implement the requirements
in subparts M and K of this part) shall
establish, implement, and maintain FFD
programs that, at a minimum, comprise
the program elements contained in this
subpart. The individuals specified in
§ 26.4(a) through (e) and (g), and, at the
licensee’s or other entity’s discretion,
§ 26.4(f), and, if necessary, § 26.4(j) shall
be subject to these FFD programs.
Licensees and other entities may rely on
the FFD program or program elements of
a C/V, as defined in § 26.5, if the C/V’s
FFD program or program elements
satisfy the applicable requirements of
this part.
■ 83. Revise § 26.51 to read as follows:
§ 26.51

Applicability.

The requirements in this subpart
apply to the licensees and other entities
identified in § 26.3(a), (b), and, as
applicable, (c) for the categories of
individuals in § 26.4(a) through (d), and,
at the licensee’s or other entity’s
discretion, in § 26.4(g) and, if necessary,
§ 26.4(j). The requirements in this
subpart also apply to the licensees and
other entities specified in § 26.3(c), as
applicable, for the categories of
individuals in § 26.4(e). At the
discretion of a licensee or other entity
in § 26.3(c), the requirements of this
subpart also may be applied to the
categories of individuals identified in
§ 26.4(f). In addition, the requirements
in this subpart apply to the entities in
§ 26.3(d) to the extent that a licensee or
other entity relies on the C/V to satisfy
the requirements of this subpart. Certain
requirements in this subpart also apply

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to the individuals specified in § 26.4(h).
The requirements in this subpart apply
to the FFD programs of licensees and
other entities identified in § 26.3(f) that
elect not to implement the requirements
in subpart M for the categories of
individuals in § 26.4 and those licensees
and other entities that elect to
implement the requirements in § 26.605.
§ 26.53

[Amended]

84. Amend § 26.53 by:
a. In paragraph (e), wherever it
appears, remove the phrase ‘‘§ 26.3(a)
through (c)’’ and add in its place the
phrase ‘‘§ 26.3(a) through (c) and (f)’’;
and
■ b. In paragraphs (g), (h), and (i),
wherever it appears, remove the phrase
‘‘(c) and (d)’’ and add in its place the
phrase ‘‘(c), (d), and (f)’’.
■
■

§ 26.63

[Amended]

85. In § 26.63, in paragraph (d) remove
the phrase ‘‘§ 26.3(a) through (d)’’ and
add in its place the phrase ‘‘§ 26.3(a)
through (d) and (f)’’.
■ 86. Revise § 26.73 to read as follows:
■

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§ 26.73

Applicability.

The requirements in this subpart
apply to the licensees and other entities
identified in § 26.3(a), (b), and, as
applicable, (c) for the categories of
individuals specified in § 26.4(a)
through (d) and (g). The requirements in
this subpart also apply to the licensees
and other entities specified in § 26.3(c),
as applicable, for the categories of
individuals in § 26.4(e). At the
discretion of a licensee or other entity
in § 26.3(c), the requirements of this
subpart also may be applied to the
categories of individuals identified in
§ 26.4(f). In addition, the requirements
in this subpart apply to the entities in
§ 26.3(d) to the extent that a licensee or
other entity relies on the C/V to satisfy
the requirements of this subpart. The
regulations in this subpart also apply to
the individuals specified in § 26.4(h)
and (j), as appropriate. The
requirements in this subpart apply to
the FFD programs of licensees and other
entities identified in § 26.3(f) that elect
not to implement the requirements in
subpart M for the categories of
individuals in § 26.4 and those licensees
and other entities that elect to
implement the requirements in
§ 26.605(b).
■ 87. Revise § 26.81 to read as follows:
§ 26.81

Purpose and applicability.

This subpart contains requirements
for collecting specimens for drug testing
and conducting alcohol tests by or on
behalf of the licensees and other entities
in § 26.3(a) through (d) for the categories

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of individuals specified in § 26.4(a)
through (d) and (g). At the discretion of
a licensee or other entity in § 26.3(c),
specimen collections and alcohol tests
must be conducted either under this
subpart for the individuals specified in
§ 26.4(e) and (f) or the licensee or other
entity may rely on specimen collections
and alcohol tests conducted under the
requirements of 49 CFR part 40 for the
individuals specified in § 26.4(e) and (f).
The requirements of this subpart do not
apply to specimen collections and
alcohol tests that are conducted under
the requirements of 49 CFR part 40, as
permitted in this paragraph and under
§§ 26.4(j) and 26.31(b)(2) and subpart K.
The requirements in this subpart apply
to the FFD programs of licensees and
other entities identified in § 26.3(f) that
elect not to implement the requirements
in subpart M for the categories of
individuals in § 26.4 and those licensees
and other entities that elect to
implement the requirements in § 26.605.
■ 88. Revise § 26.201 to read as follows:
§ 26.201

Applicability.

(a) The requirements in this subpart,
with the exception of § 26.202, apply to
the licensees and other entities
identified in § 26.3(a); if applicable, (c),
(d), and (f), for licensees and other
entities not implementing the
requirements in subparts K and M. For
the licensees and other entities to whom
the requirements in this subpart, with
the exception of § 26.202, apply, the
requirements in §§ 26.203 and 26.211
apply to the individuals identified in
§ 26.4(a) through (c). In addition, the
requirements in §§ 26.205 through
26.209 apply to the individuals
identified in § 26.4(a).
(b) The requirements in this subpart,
with the exception of § 26.203, apply to
the licensees or other entities identified
in § 26.3(f) implementing this subpart
under §§ 26.604 and 26.605. For these
licensees and other entities, the
requirements in §§ 26.202 and 26.211
apply to the individuals identified in
§ 26.4(a) through (c) and any person
licensed to operate under 10 CFR part
53; and the requirements in §§ 26.205
through 26.209 apply to the individuals
identified in § 26.4(a).
■ 89. Add § 26.202 to read as follows:
§ 26.202 General provisions for facilities
licensed under part 53.

(a) Policy. Licensees must establish a
policy for the management of fatigue for
all individuals who are subject to the
licensee’s FFD program and incorporate
it into the written policy required in
§ 26.606(a).
(b) Procedures. In addition to the
procedures required in § 26.606(b),

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licensees must develop, implement, and
maintain procedures that—
(1) Describe the process to be
followed when any individual
identified in § 26.4(a) through (c) makes
a self-declaration that he or she is not
fit to safely and competently perform
his or her duties for any part of a
working tour as a result of fatigue. The
procedure must—
(i) Describe the individual’s and
licensee’s rights and responsibilities
related to self-declaration;
(ii) Describe requirements for
establishing controls and conditions
under which an individual may be
permitted or required to perform work
after that individual declares that he or
she is not fit due to fatigue; and
(iii) Describe the process to be
followed if the individual disagrees
with the results of a fatigue assessment
that is required under § 26.211(a)(2);
(2) Describe the process for
implementing the controls required
under § 26.205 for the individuals who
are performing the duties listed in
§ 26.4(a);
(3) Describe the process to be
followed in conducting fatigue
assessments under § 26.211; and
(4) Describe the disciplinary actions
that the licensee may impose on an
individual following a fatigue
assessment, and the conditions and
considerations for taking those
disciplinary actions.
(c) Training and assessments.
Licensees must include the following
KAs in the content of the training and
trainee assessments required in
§ 26.608:
(1) Knowledge of the contributors to
worker fatigue, circadian variations in
alertness and performance, indications
and risk factors for common sleep
disorders, shiftwork strategies for
obtaining adequate rest, and the
effective use of fatigue countermeasures;
and
(2) Ability to identify symptoms of
worker fatigue and contributors to
decreased alertness in the workplace.
(d) Recordkeeping. Licensees must
retain the following records for at least
3 years or until the completion of all
related legal proceedings, whichever is
later:
(1) Records of work hours for
individuals who are subject to the work
hour controls in § 26.205;
(2) For licensees implementing the
requirements of § 26.205(d)(3), records
of shift schedules and shift cycles, or,
for licensees implementing the
requirements of § 26.205(d)(7), records
of shift schedules and records showing
the beginning and end times and dates
of all averaging periods, of individuals

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who are subject to the work hour
controls in § 26.205;
(3) The documentation of waivers that
is required in § 26.207(a)(4), including
the bases for granting the waivers;
(4) The documentation of work hour
reviews that is required in § 26.205(e)(3)
and (e)(4); and
(5) The documentation of fatigue
assessments that is required in
§ 26.211(g).
(e) Reporting. Licensees must include
the following information in a standard
format in the annual FFD program
performance report required under
§ 26.617(b)(2):
(1) A summary for each nuclear power
plant site of all instances during the
previous calendar year when the
licensee waived one or more of the work
hour controls specified in § 26.205(d)(1)
through (d)(5)(i) and (d)(7) for
individuals described in § 26.4(a). The
summary must include only those
waivers under which work was
performed. If it was necessary to waive
more than one work hour control during
any single extended work period, the
summary of instances must include
each of the work hour controls that were
waived during the period. For each
category of individuals specified in
§ 26.4(a), the licensee must report—
(i) The number of instances when
each applicable work hour control
specified in § 26.205(d)(1)(i) through
(iii), (d)(2)(i) and (ii), (d)(3)(i) through
(v), and (d)(7) was waived for
individuals not working on outage
activities;
(ii) The number of instances when
each applicable work hour control
specified in § 26.205(d)(1)(i) through
(iii), (d)(2)(i) and (ii), (d)(3)(i) through
(v), (d)(4) and (d)(5)(i), and (d)(7) was
waived for individuals working on
outage activities; and
(iii) A summary that shows the
distribution of waiver use among the
individuals applicable within each
category of individuals identified in
§ 26.4(a) (e.g., a table that shows the
number of individuals who received
only one waiver during the reporting
period, the number of individuals who
received a total of two waivers during
the reporting period).
(2) A summary of corrective actions,
if any, resulting from the analyses of
these data, including fatigue
assessments.
(f) Audits. Licensees must audit the
management of worker fatigue under
§ 26.615.
■ 90. In § 26.205, revise paragraphs
(d)(7)(iii) and (d)(8) to read as follows:
§ 26.205

*

*

Work Hours.

*

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*

*

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(d) * * *
(7) * * *
(iii) Each licensee shall state, in its
FFD policy and procedures required by
either §§ 26.27 and 26.203(a) and (b) or
§§ 26.202(a) and (b) and 26.606, the
work hour counting system in
§ 26.205(d)(7)(ii) the licensee is using.
(8) Each licensee shall state, in its
FFD policy and procedures required by
either §§ 26.27 and 26.203(a) and (b) or
§§ 26.202(a) and (b) and 26.606, the
requirements with which the licensee is
complying: the minimum days off
requirements in § 26.205(d)(3) or
maximum average work hours
requirements in § 26.205(d)(7).
*
*
*
*
*
■ 91. In § 26.207, revise paragraph
(a)(1)(ii) to read as follows:
§ 26.207

Waivers and exceptions.

(a) * * *
(1) * * *
(ii) A supervisor assesses the
individual face to face and determines
that there is reasonable assurance that
the individual will be able to safely and
competently perform his or her duties
during the additional work period for
which the waiver will be granted. The
supervisor performing the assessment
shall be trained as required by either
§§ 26.29 and 26.203(c) or §§ 26.202(c)
and 26.608 and shall be qualified to
direct the work to be performed by the
individual. If there is no supervisor on
site who is qualified to direct the work,
the assessment may be performed by a
supervisor who is qualified to provide
oversight of the work to be performed by
the individual. At a minimum, the
assessment must address the potential
for acute and cumulative fatigue
considering the individual’s work
history for at least the past 14 days, the
potential for circadian degradations in
alertness and performance considering
the time of day for which the waiver
will be granted, the potential for fatiguerelated degradations in alertness and
performance to affect risk-significant
functions, and whether any controls and
conditions must be established under
which the individual will be permitted
to perform work. For licensees and other
entities in § 26.3(f), the assessment may
be performed remotely using electronic
communications. In such instances, the
assessment must be supported by
someone who is present in-person with
the individual whose alertness may be
impaired, and that supporting person
must be trained under the requirements
of either § 26.29 and § 26.203(c) or
§ 26.202(c) and § 26.608.
*
*
*
*
*

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92. In § 26.211, revise paragraphs
(a)(1) and (3) and paragraph (b)
introductory text to read as follows:

■

§ 26.211

Fatigue assessments.

(a) * * *
(1) For-cause. In addition to any other
test or determination of fitness that may
be required under §§ 26.31(c), 26.77,
26.607(b), and 26.619, a fatigue
assessment must be conducted in
response to an observed condition of
impaired individual alertness creating a
reasonable suspicion that an individual
is not fit to safely and competently
perform his or her duties, except if the
condition is observed during an
individual’s break period. If the
observed condition is impaired alertness
with no other behaviors or physical
conditions creating a reasonable
suspicion of possible substance abuse,
then the licensee need only conduct a
fatigue assessment. If the licensee has
reason to believe that the observed
condition is not due to fatigue, the
licensee need not conduct a fatigue
assessment;
*
*
*
*
*
(3) Post-event. A fatigue assessment
must be conducted in response to events
requiring post-event drug and alcohol
testing as specified in § 26.31(c) or postevent tests in § 26.607(b)(4). Licensees
may not delay necessary medical
treatment in order to conduct a fatigue
assessment; and
*
*
*
*
*
(b) Only supervisors and FFD program
personnel who are trained under either
§§ 26.29 and 26.203(c) or §§ 26.202(c)
and 26.608 may conduct a fatigue
assessment. The fatigue assessment
must be conducted face to face with the
individual whose alertness may be
impaired. For licensees and other
entities in § 26.3(f), a fatigue assessment
may be performed remotely using
electronic communications. In such
instances, the fatigue assessment must
be supported by someone who is
present in-person with the individual
whose alertness may be impaired, and
that supporting person must be trained
in accordance with the requirements of
either §§ 26.29 and 26.203(c) or
§§ 26.202(c) and 26.608.
*
*
*
*
*
■ 93. Add Subpart M, consisting of
§§ 26.601 through 26.619, to read as
follows:
Subpart M—Fitness for Duty Programs
for Facilities Licensed Under 10 CFR
Part 53
Sec.
26.601
26.603

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26.604 FFD program requirements for
facilities that satisfy the § 26.603(c)
criterion.
26.605 FFD program requirements for
facilities that do not implement § 26.604.
26.606 Written policy and procedures.
26.607 Drug and alcohol testing.
26.608 FFD program training.
26.609 Behavioral observation.
26.610 Sanctions.
26.611 Protection of information.
26.613 Appeals process.
26.615 Audits.
26.617 Recordkeeping and reporting.
26.619 Suitability and fitness
determinations.
§ 26.601

Applicability.

A licensee or other entity in § 26.3(f),
at its discretion, may establish,
implement, and maintain a fitness-forduty (FFD) program that satisfies the
requirements of this subpart for those
categories of individuals in § 26.4, as
applicable, and any person licensed to
operate under 10 CFR part 53. If a
licensee or other entity in § 26.3(f) does
not elect to implement an FFD program
that satisfies the requirements of this
subpart, then those categories of
individuals in § 26.4, as applicable, and
any person licensed to operate under 10
CFR part 53 must be subject to an FFD
program that satisfies all part 26
requirements, except for those
requirements in subparts K and M.

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§ 26.603

General provisions.

(a) FFD program description. An
applicant’s description of the FFD
program in its Final Safety Analysis
Report, required by subpart H of part 53
of this chapter, must include—
(1) If the applicant performed the
analysis under paragraph (c) of this
section, a summary of the analysis,
including the assumptions,
methodology, conclusion, and
references;
(2) A statement whether the FFD
program will be implemented pursuant
to § 26.604 or § 26.605, or will satisfy all
part 26 requirements, except for the
requirements in subparts K and M;
(3) A discussion of the applicability of
the FFD program to those individuals
described in § 26.4 and how the
program will be implemented offsite at
a U.S. Nuclear Regulatory Commission
(NRC)-licensed facility authorized to
assemble or test a manufactured reactor,
if applicable;
(4) A description of the drug and
alcohol testing and fitness
determination process to be
implemented through the licensee’s or
other entity’s procedures, including the
collection and testing facilities to be
used, biological specimens to be
collected, and sanctions to be imposed

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upon a confirmed FFD policy violation;
and
(5) A summary of the FFD
performance monitoring and review
program (PMRP), including the
measures and thresholds required by
paragraph (d)(1) of this section.
(b) FFD program implementation and
availability. For the licensees and other
entities in § 26.3(f), other than the
holder of a manufacturing license (ML),
the FFD program must be implemented
no later than the start of construction
activities, as defined in § 26.5, and
maintained until the NRC’s docketing of
the license holder’s certifications
described in § 53.1070 of this chapter.
For holders of an ML, the FFD program
must be implemented no later than the
start of activities that assemble the
manufactured reactor and maintained
until expiration of the ML.
(c) Criterion and analysis for an FFD
program. For a licensee or other entity
to implement an FFD program under
§ 26.604, the licensee or other entity
must perform a site-specific analysis to
demonstrate that the facility and its
operation satisfy the criterion in
§ 53.860(a)(2) of this chapter. The
licensee or other entity must maintain
the analysis, including updates to reflect
changes made to the staffing, FFD
programs, or offsite support resources
described in the analysis, to show that
the facility and its operation continue to
satisfy the criterion, until permanent
cessation of operations under § 53.1070
of this chapter.
(d) FFD performance monitoring and
review. A licensee or other entity must
establish performance measures and
associated thresholds as described in
paragraph (d)(1) of this section and
monitor the effectiveness of its FFD
program by comparing performance data
against these performance measures and
thresholds, in a manner sufficient to
satisfy the § 26.23 performance
objectives.
(1) PRMP elements. The PMRP must
be documented and maintained and
include the following program elements:
(i) Performance measures.
Performance measures must be
identified and designed to monitor FFD
program performance.
(A) If the licensee or other entity is
subject to the requirements in § 26.604,
then the monitoring program must
include performance measures for the
following: the behavioral observation
program; occurrence of FFD policy
violations categorized by licensee
employee, contractor/vendor, and labor
category; and occurrence of individuals
with potentially disqualifying
information or who possessed FFD
prohibited items.

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87031

(B) If the licensee or other entity is
subject to the requirements in § 26.604
and has implemented a drug testing
program at its discretion or is subject to
the requirements of § 26.605, then the
monitoring program must include
performance measures identified in
paragraph (d)(1)(i)(A) of this section.
This monitoring program must also
include performance measures for the
pre-access and random positive testing
rates, random testing rate for licensee
employees and contractor/vendors, and
the number of subversion attempts
categorized by licensee employee,
contractor/vendor, and labor category.
(ii) Thresholds. Licensee- or other
entity-specific thresholds for its sitespecific performance measures must be
established and used to facilitate
corrective actions to maintain FFD
program performance. Initial thresholds
must be based on FFD performance data
from comparable facilities subject to
part 26, the licensee’s or other entity’s
fleet-level program performance if
applicable, and industry FFD
performance data.
(iii) Monitoring program. Licensees
and other entities must monitor the
performance of their FFD programs
against licensee- or other entityestablished performance measures and
thresholds as FFD performance data is
received to determine whether a
threshold has been exceeded. Licensees
and other entities must perform year-toyear comparisons of site-specific
performance; site-specific performance
to the licensee’s or other entity’s fleetlevel program performance, if
applicable; and site-specific to industry
performance.
(iv) Quantitative and qualitative
reviews. The PMRP must include a
documented review of the elements in
paragraph (d)(1)(i) through (iii) of this
section and the following qualitative
elements.
(A) Worker protections. The review
must include a documented assessment
of the licensee’s or other entity’s
implementation of the protections
described in §§ 26.606(b)(1)(iii), 26.611,
and 26.613.
(B) Laboratory test results and
Medical Review Officer performance.
The review must include a documented
assessment of whether the actions taken
by the Medical Review Officer (MRO)
met the requirements in § 26.185 based
on the laboratory test results reported
under § 26.169. This review must
include a comparative analysis between
the point of collection testing and
assessment (POCTA) screening result(s)
and the corresponding specimen test
results obtained from the U.S.
Department of Health and Human

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Services (HHS)-certified laboratory if
the POCTA indicated a positive,
adulterated, substituted, or invalid
screening result or discrepant biological
marker, to assess the effectiveness of the
POCTA and to inform MRO decisions
under § 26.185 or § 26.607(m)(6).
(C) Change control. The review must
include a documented assessment of the
changes made under paragraph (e) of
this section to verify that the summation
of program changes has not resulted in
a reduction in FFD program
effectiveness.
(2) Corrective actions. Corrective
actions must be implemented to address
when FFD performance meets a
licensee-established performance
threshold or to resolve a finding
resulting from a qualitative review or
audit in a manner that restores
performance and corrects root causes,
contributing causes, or both.
(3) Program review periodicity. The
documented review in paragraph
(d)(1)(iv) of this section must be
conducted biennially to assess and
modify licensee or other entity
implementation of its FFD program.
This documented review must
demonstrate that the performance
measures and thresholds are appropriate
and adjusted as necessary based on sitelevel and licensee’s or other entity’s
fleet-level, if applicable, program
performance, and industry performance.
(i) Identified program weaknesses and
corrective actions must be summarized
in the annual reporting requirement
described in § 26.617(b)(2) or § 26.717,
as applicable.
(ii) The program review must be
completed and approved by the licensee
or other entity to support the reporting
of PMRP weaknesses and corrective
actions as required in paragraph (d)(3)(i)
of this section every odd-numbered
year, and the implementation of
corrective actions before May 15 of that
odd-numbered year.
(e) FFD program change control. (1)
The licensee or other entity may make
changes to its FFD program under this
subpart if—
(i) The licensee or other entity
performs and retains an analysis
demonstrating that the changes do not
reduce the effectiveness of the FFD
program; or
(ii) The change was necessitated or
justified by a change to part 26,
laboratory processes or procedures, or
guidance issued by the HHS or NRC, as
implemented by the licensee or other
entity though its procedures.
(2) A licensee or other entity desiring
to make a change that decreases FFD
program effectiveness must implement a
mitigating strategy so the FFD program,

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as revised, will continue to satisfy the
performance objectives in § 26.23 and
not result in a reduction in program
effectiveness.
(3) Except for phencyclidine, and
notwithstanding paragraph (e)(1)(ii) of
this section, the change control process
may not be used to reduce the minimum
panel of drugs to be tested in
§ 26.607(c)(1).
(4) The licensee must retain a record
of each change made under this section
for a period of at least 5 years from the
date the change was implemented and
summarize this change in its annual
FFD performance report required by
§ 26.617(b)(2) or § 26.717, as applicable.
§ 26.604 FFD program requirements for
facilities that satisfy the § 26.603(c)
criterion.

(a) FFD program. A licensee or other
entity with an analysis that
demonstrates that its facility and
operation satisfy the criterion in
§ 26.603(c) may elect to establish,
implement, and maintain an FFD
program under this section. That FFD
program must contain the following
elements:
(1) Applies to those individuals
described in § 26.4, as applicable; and
(2) Implements the following
requirements and subparts in this part:
(i) § 26.23, Performance objectives;
(ii) § 26.603, General provisions;
(iii) § 26.606, Written policies and
procedures, (a) and, if applicable (b);
(iv) § 26.608, FFD program training;
(v) § 26.609, Behavioral observation;
(vi) § 26.610, Sanctions;
(vii) § 26.611, Protection of
information;
(viii) § 26.613, Appeals process;
(ix) § 26.615, Audits;
(x) § 26.617, Recordkeeping and
reporting;
(xi) § 26.619, Suitability and fitness
determinations;
(xii) Subpart A—Administrative
Provisions;
(xiii) Subpart I—Managing Fatigue;
and
(xiv) Subpart O—Inspections,
Violations, and Penalties.
(b) [Reserved]
§ 26.605 FFD program requirements for
facilities that do not implement § 26.604.

(a) Licensees and other entities who
satisfy the criterion in § 26.603(c), at
their discretion, and licensees and other
entities who do not satisfy the criterion
in § 26.603(c), must establish,
implement, and maintain an FFD
program under this section either during
construction activities as defined in
§ 26.5, or during activities performed
under an ML that allows the assembly,

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testing, or both of a manufactured
reactor, as applicable. This FFD program
must contain the following elements:
(1) Applies to those individuals
described in § 26.4, as applicable; and,
(2) Implements the following
requirements and subparts in this part—
(i) § 26.23, Performance objectives;
(ii) § 26.603, General provisions;
(iii) § 26.606, Written policy and
procedures;
(iv) § 26.607, Drug and alcohol testing;
(v) § 26.608, FFD program training;
(vi) § 26.609, Behavioral observation;
(vii) § 26.610, Sanctions;
(viii) § 26.611, Protection of
information;
(ix) § 26.613, Appeals process;
(x) § 26.615, Audits;
(xi) § 26.617, Recordkeeping and
reporting;
(xii) § 26.619, Suitability and fitness
determinations;
(xiii) Subpart A—Administrative
Provisions;
(xiv) Subpart I—Managing Fatigue, in
the case of holders of an ML that allows
the assembly, testing, or both of a
manufactured reactor; and
(xv) Subpart O—Inspections,
Violations, and Penalties.
(b) Licensees and other entities who
satisfy the criterion in § 26.603(c), at
their discretion, and licensees and other
entities who do not satisfy the criterion
in § 26.603(c), before the loading of fuel
onsite into a reactor vessel; before
receiving a manufactured reactor; or
before individuals subject to part 26
operate, test, perform maintenance of, or
direct the maintenance or surveillance
of security-related equipment or
equipment that a risk-informed
evaluation process has shown to be
significant to public health and safety,
must establish, implement, and
maintain an FFD program that—
(1) Applies to those individuals
described in § 26.4, as applicable; and,
(2) Implements the following
requirements and subparts—
(i) § 26.23, Performance objectives;
(ii) § 26.603, General provisions;
(iii) § 26.606, Written policy and
procedures;
(iv) § 26.607, Drug and alcohol testing;
(v) § 26.608, FFD program training;
(vi) § 26.609, Behavioral observation;
(vii) § 26.611, Protection of
information;
(viii) § 26.613, Appeals process;
(ix) § 26.615, Audits;
(x) Subpart A—Administrative
Provisions;
(xi) Subpart C—Granting and
Maintaining Authorization;
(xii) Subpart D—Management Actions
and Sanctions to be Imposed;
(xiii) Subpart H—Determining
Fitness-for-Duty Policy Violations and

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Determining Fitness, unless using the
HHS Guidelines for MRO evaluation of
drug test results, and determining
fitness;
(xiv) Subpart I—Managing Fatigue;
(xv) Subpart N—Recordkeeping and
Reporting Requirements; and
(xvi) Subpart O—Inspections,
Violations, and Penalties.

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§ 26.606

Written policy and procedures.

(a) Licensees and other entities that
implement an FFD program under this
subpart must ensure that—
(1) A written FFD policy statement is
provided to each individual who is
subject to the program before the
individual is subject to behavioral
observation, drug and alcohol testing, or
both.
(2) The FFD policy statement
describes the performance objectives in
§ 26.23.
(3) The FFD policy statement
describes the minimum days off
requirements in § 26.205(d)(3) or
maximum average work hours
requirements in § 26.205(d)(7).
(4) The FFD policy statement must be
written in sufficient detail to provide
affected individuals with information
on what is expected of them and what
consequences may result from a lack of
adherence to the policy, including those
elements described in § 26.606(b), part
26-required sanctions, and required
medical/clinical treatment and followup testing for FFD policy violations.
(5) The FFD policy statement
describes the individual’s
responsibilities to report for work in a
physiological and psychological
condition that enables the safe and
competent performance of assigned
duties and responsibilities and inform a
licensee- or other entity-designated
representative when the individual
determines that this cannot be
accomplished.
(b) Licensees and other entities must
establish, implement, and maintain
written procedures that address the
following topics:
(1) If implementing a drug and
alcohol testing program under this
subpart,
(i) The methods and techniques to
collect and test for drugs and alcohol
and for the shipping and temporary
storage of biological specimens used for
drug testing at HHS-certified
laboratories,
(ii) The urine specimen volumes,
techniques for split specimen
collections, and the acceptability of a
urine specimen as described in § 26.111
or as described in the HHS Guidelines,
(iii) Protecting the privacy of an
individual who provides a specimen,

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protecting the integrity of the specimen,
and ensuring that the test results are
valid and attributable to the correct
individual, and
(iv) If the licensee or other entity
elects to use the HHS Guidelines, the
name of the specific HHS Guideline and
revision being implemented by the
licensee or other entity and a
description of the specific sections in
the guideline that are being
implemented in the procedure,
including specimen collections, drug
testing, and evaluation of test results.
(2) The immediate and follow-up
actions that will be taken, and the
procedures to be used, in those cases in
which individuals who are subject to
the FFD program:
(i) Have been involved in the use,
sale, or possession of illegal substances,
illegal drugs, or illicit substances;
(ii) Are impaired by any illegal
substances, illegal drugs, or illicit
substances or the consumption of
alcohol as determined by behavioral
observation or a test that measures
blood alcohol concentration;
(iii) If drug and alcohol testing is
conducted, attempted to subvert the
testing process by adulterating or
diluting specimens (in vivo or in vitro),
substituting specimens, or by any other
means;
(iv) If drug and alcohol testing is
conducted, refused to provide a
specimen for analysis or follow
instructions provided by FFD program
personnel;
(v) Had legal action taken relating to
drug or alcohol use; or
(vi) Demonstrated character or actions
indicating that the individual cannot be
trusted or relied upon to perform those
duties and responsibilities or maintain
access to NRC-licensed facilities, special
nuclear material (SNM), or sensitive
information.
(3) The process, including the duties
and responsibilities of FFD program
personnel, to be followed if an
individual’s behavior or condition raises
a concern regarding the possible use,
sale, or possession of illegal drugs onor offsite; the possible use or possession
of alcohol on the NRC-licensed facility;
impairment from any cause that in any
way could adversely affect the
individual’s ability to safely and
competently perform the individual’s
duties; or the receipt of credible
information indicating that the
individual cannot be trusted or relied on
to perform those duties and
responsibilities making the individual
subject to this part.
(4) Operation and oversight of an
onsite or offsite collection facility.

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(5) The fatigue management
requirements in §§ 26.202(b) and either
26.205(d)(3) or (d)(7).
(6) Measures to prevent subversion of
drug and alcohol tests conducted onsite
and offsite.
§ 26.607

Drug and alcohol testing.

Licensees and other entities
implementing § 26.604, at their
discretion, and licensees and other
entities implementing § 26.605 must
perform drug and alcohol testing that
complies with the following
requirements—
(a) Split specimens. Split specimen
collections of oral fluid or urine must be
used for the test conditions described in
paragraph (b) of this section. A split
specimen collection need not be used if
the licensee or other entity elects to use
a POCTA device for a screening test
conducted during random testing under
paragraphs (b)(2) and (h) of this section
or a protected area portal monitor
indication that drugs or alcohol were
detected under paragraph (j) of this
section. Testing of the split specimen
(specimen B) requires the donor’s
permission unless ordered by the MRO
to resolve an invalid test result obtained
for specimen A.
(b) Test conditions. Individuals
identified in § 26.4 must be subject to
drug and alcohol testing under the
following conditions:
(1) Pre-access. A pre-access test must
be conducted for drugs and alcohol
before performing or directing the
conduct of roles and responsibilities
making the individual subject to this
subpart or being granted unescorted
access to the protected areas of the NRClicensed facility. A pre-access test must
have been conducted no more than 14
days before the individual is granted
unescorted access.
(2) Random. Random testing for drugs
and alcohol must—
(i) Be administered in a manner that
provides reasonable assurance that
individuals are unable to predict the
time periods during which specimens
will be collected;
(ii) Require individuals who are
selected for random testing to report to
the onsite collection site as soon as
reasonably practicable after notification,
within the time period specified in the
FFD program procedure;
(iii) Ensure that all individuals in the
population that is subject to random
testing on a given day have an equal
probability of being selected and tested;
(iv) Ensure that an individual
completing a test is immediately eligible
for another random test; and
(v) Ensure that the sampling process
used to select individuals for random

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testing provides that the number of
random tests performed annually is
equal to at least 50 percent for licensee
employees and 50 percent for
contractor/vendors at the NRC-licensed
site.
(3) For-cause. A for-cause drug test,
alcohol test, or both, must be conducted
onsite in response to an individual’s
observed behavior or physical condition
indicating possible substance abuse or
after receiving credible information that
an individual is engaging in substance
abuse, as defined in § 26.5;
(4) Post-event. A post-event test for
drugs and alcohol must be conducted—
(i) As soon as practical after an event
involving a human error that was
committed by an individual specified in
§ 26.4, where the human error may have
caused or contributed to the event. This
test must be conducted onsite unless the
individual requires offsite medical care.
The licensee or other entity must test
the individual(s) who committed or
directed the error and need not test
individuals who were affected by the
event and whose actions likely did not
cause or contribute to the event. The
licensee or other entity must describe in
its procedures what constitutes a human
error.
(ii) Within 4 hours of an event unless
immediate medical intervention
precludes the conduct of the test on the
individual(s) who caused or contributed
to the accident(s), if the event results
in—
(A) An illness or personal injury to
any individual which results in death,
days away from work, restricted work,
transfer to another job, medical
treatment beyond first aid, loss of
consciousness, or other significant
illness or injury, as diagnosed by a
licensee- or other entity-designated
physician or other licensed health care
professional, even if the illness or injury
does not result in death, days away from
work, restricted work or job transfer,
medical treatment beyond first aid, or
loss of consciousness; or
(B) Damage to any safety- or securityrelated structures, systems, and
components; and
(5) Follow-up. An individual subject
to part 26 who has violated the FFD
policy for substance use or abuse, or the
sale, use, or possession of illegal drugs
must be subject to a follow-up series of
tests for drugs, alcohol, or both to verify
an individual’s continued abstinence
from substance abuse.
(c) Urine and oral fluid specimens. (1)
All urine or oral fluid specimens must
be subject to validity testing, including
an adulterant and biological marker, and
tested for the substances listed in

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§ 26.31(d)(1), except as allowed by
§ 26.603(e)(3).
(2) For the use of urine as the
biological specimen to be tested, the
following requirements must be
implemented—
(i) § 26.115, Collecting a urine
specimen under direct observation;
(ii) § 26.119, Determining ‘‘shy’’
bladder; and
(iii) § 26.163, Cutoff levels for drugs
and drug metabolites, (a)(2) regarding
special analysis testing.
(3) For alcohol testing onsite, the
following requirements must be
implemented—
(i) § 26.91, Acceptable devices for
conducting initial and confirmatory
tests for alcohol and methods of use;
(ii) § 26.93, Preparing for alcohol
testing;
(iii) § 26.95, Conducting an initial test
for alcohol using a breath specimen;
(iv) § 26.97, Collecting oral fluid
specimens for alcohol and drug testing;
(v) § 26.99, Determining the need for
a confirmatory test for alcohol;
(vi) § 26.101, Conducting a
confirmatory test for alcohol; and,
(vii) § 26.103, Determining a
confirmed positive test result for
alcohol.
(4) For all test conditions in paragraph
(b) of this section, except for the use of
a POCTA screening device in paragraph
(h) of this section, and for MRO-directed
tests under § 26.185, drug testing must
be performed at an HHS-certified
laboratory for the specific biological
specimen to be tested. Only HHScertified laboratory test results from
urine and oral fluid specimens may be
used for the issuance of a part 26required sanction. The licensee or other
entity must establish and maintain a
contract with a primary and a back-up
HHS-certified laboratory (with a
different Certifying Scientist) for the
specimen(s) to be tested. These
contracts must stipulate that the
laboratories are subject to inspection or
audit by the licensee or other entity and
that records and documents must be
provided and/or able to be photocopied
and removed from the premises to
support the inspection or audit.
(d) Privacy and integrity. The
specimen collection and drug and
alcohol testing procedures of FFD
programs must protect the donor’s
privacy and the integrity of the
specimen and implement quality
controls to ensure that test results are
valid and attributable to the correct
individual.
(e) Offsite collection facilities. At the
licensee’s or other entity’s discretion,
specimen collections and alcohol testing
may be conducted at a local hospital or

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other facility licensed to conduct
specimen collections and perform
alcohol testing and audited by the State
or a State-designated entity. The
licensee or other entity must audit these
facilities, if used, before their initial use
and then on a biennial basis to confirm
that the facility procedures are
comparable to those described in
subpart E of this part or the HHS
Guidelines for urine and oral fluid.
(f) Initial testing. A licensee or other
entity subject to this subpart performing
an initial test must use an
immunoassay, or an alternative
technology established in its FFD
program through § 26.603(e), that
satisfies the requirements of the U.S.
Food and Drug Administration (FDA)
for commercial distribution. Specimens
that yield positive, adulterated,
substituted, or invalid initial validity or
drug test results or discrepant biological
markers must be subject to confirmatory
testing by an HHS-certified laboratory,
certified for that biological specimen,
except for invalid specimens that cannot
be tested.
(g) Oral fluid testing. If the licensee or
other entity elects to use oral fluid for
drug or alcohol testing, the collection,
packaging, and temporary storage of the
drug or alcohol test device, and
shipment of an oral fluid specimen to an
HHS-certified laboratory or the
collection of an oral fluid specimen for
alcohol testing must be performed in
accordance with licensee- or other
entity-established procedures based
either on the requirements in part 26 or
the procedures in HHS Guidelines
identified by the licensee or other entity
in § 26.606(b)(1)(iv). The device must
have received premarket approval from
the FDA and must not expire before
laboratory testing. The drugs, drug
metabolites, initial and confirmatory
testing cutoffs, and biological markers, if
applicable, must be those established by
HHS for oral fluid testing and the
alcohol cutoffs in this part or, if not
established by HHS or the NRC for the
panel of drugs and drug metabolites to
be tested, as determined and
documented by a forensic toxicologist
review conducted pursuant to
§ 26.31(d)(1)(i)(D).
(h) Point of collection testing and
assessment. (1) If the licensee or other
entity elects to use a POCTA device,
then it may only be used for pre-access
and random drug and alcohol initial
testing in paragraph (b) of this section,
the alcohol testing process in paragraph
(c)(3) of this section, and the portal area
screening process in paragraph (j) of this
section. Before the licensee or other
entity uses a POCTA device, a forensic
toxicologist must review and document

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their evaluation that the validity and
accuracy of the device for alcohol and/
or the drugs and drug metabolites listed
in § 26.31(d) are comparable to the
performance achieved by initial testing
conducted using a similar technology at
an HHS-certified laboratory. For initial
testing of drugs and drug metabolites
using a POCTA device, this review must
include a documented evaluation of
POCTA device performance against the
requirements in § 26.161(b) for a urine
specimen or the procedures in the HHS
Guidelines for urine or oral fluid, as
implemented by the licensee or other
entity through its procedures.
(2) If the performance of the POCTA
device is not comparable to that
achieved from initial testing conducted
by an HHS-certified laboratory as
determined by the forensic toxicologist,
then the licensee or other entity must
implement a mitigating strategy to
maintain program effectiveness under
§ 26.603(e)(2), as applicable.
(3) The licensee and other entity must
implement procedures for the use of a
POCTA that ensures the effectiveness of
the collection process, assessment of the
screening results, and prevention of
subversion attempts.
(4) If the use of a POCTA device
indicates a discrepant biological marker
or that a test result exceeds the initial
test cutoff, the specimen is invalid, or
the individual subverted the drug or
alcohol test, then the individual must be
immediately removed from duties,
responsibilities, and access making the
individual subject to this subpart.
(i) The individual must be subject to
an immediate drug and alcohol test
using the alcohol testing process in
paragraph (c)(3) of this section for a
positive alcohol screen and either oral
fluid or urine by a collection kit that is
not a POCTA device, but of the same
type of biological specimen collected by
the POCTA, for validity, if required, and
initial and confirmatory testing by an
HHS-certified laboratory.
(ii) If this individual shows any signs
of impairment, the individual’s
authorization must be temporarily
removed until the MRO reviews the
laboratory test result(s), interviews the
individual, and performs a
determination of fitness under § 26.189
or § 26.619, as applicable, that enables
the restoration of authorization.
(i) Hair testing. The testing of hair
specimens may only be used to inform
a licensee’s or other entity’s
determination of whether the individual
is trustworthy and reliable under the
test condition in paragraph (b)(1) of this
section to supplement the information
gained from a pre-access test using oral
fluid or urine as the test specimen and

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must be conducted at an HHS-certified
laboratory certified for hair specimens.
(1) If used, this process must be
described in the licensee’s or other
entity’s FFD policy and described in
detail in its procedure. The panel of
drugs and drug metabolites to be
evaluated must only include those listed
as Schedule I or II of section 202 of the
Controlled Substances Act [21 U.S.C.
812]. The collection, packaging, and
temporary storage of a hair specimen
and shipment of the specimen to an
HHS-certified laboratory must be
conducted in accordance with the HHS
Guidelines. The test kit must be FDA
cleared, and licensee- or other entitydesignated FFD program personnel must
conduct the collection, packaging,
temporary storage, shipping, and
custody and control of the specimen.
(2) Before the licensee or other entity
begins to conduct hair testing, the initial
and confirmatory testing cutoffs must be
the cutoffs established by HHS for hair
testing or, if not established by HHS or
the NRC, as determined by a forensic
toxicologist review conducted pursuant
to § 26.31(d)(1)(i)(D).
(3) Confirmed positive test results
must be considered potentially
disqualifying FFD information until
proven otherwise by a review under
§ 26.613. Sanctions under this subpart
must not be issued for any FFD policy
violation involving a drug test using a
hair specimen unless the licensee or
other entity determines that the
individual subverted, as defined in
§ 26.5, the hair test.
(j) Portal area screening. A noninvasive point of collection testing
instrument may be used to screen
individuals for drugs, drug metabolites,
and alcohol before the individuals’
entry into or exit from a protected or
vital area.
(1) If a licensee or other entity uses
such an instrument, then before such
use, a forensic toxicologist must review
the instrument and document an
evaluation that the instrument and
setpoints used in the instrument are
acceptable for use for the detection and
screening of the drugs and drug
metabolites selected for screening from
the panel of drugs and drug metabolites
to be tested under the FFD program and
alcohol and its metabolites.
(2) The instrument must be operated
in accordance with the manufacturer’s
specifications. If screening detects the
presence of drugs, drug metabolites, or
alcohol at or above the instrument set
point(s), the individual screened by the
instrument must be subject to a POCTA
screening test using the process
described in paragraph (h) of this

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87035

section or an oral fluid or urine test that
is sent to an HHS-certified laboratory.
(3) A part 26 sanction may not be
issued to an individual based solely on
a portal area screening instrument
detection that drugs or alcohol exceed
the instrument’s established setpoint.
(k) Blood testing. The testing of blood
specimens may only be conducted
under the order of the licensee- or other
entity-designated MRO for a valid
medical reason as confirmed by the
MRO pursuant to § 26.31(d)(5). This
specimen must be subject to testing by
a laboratory that satisfies quality control
requirements that are comparable to
those required for certification by the
HHS.
(l) Custody-and-control form. For the
collection and packaging of urine, oral
fluid, and hair specimens, the licensee
or other entity must use a custody-andcontrol form approved by the U.S.
Office of Management and Budget. For
the use of a POCTA device, the licensee
or other entity must implement a
licensee- or other entity-approved and
-maintained procedure that ensures the
reliability of the tracking, handling, and
storage of a specimen from the point of
specimen collection to the final
disposition of the specimen and the
reliability of an identification system to
uniquely assign the specimen to the
donor.
(m) Medical Review Officer. Licensees
or other entities must—
(1) Require their designated MRO to
review positive, adulterated,
substituted, and dilute confirmatory
drug and validity test results and test
results of questionable validity to
determine whether the donor has
violated the FFD policy for urine and
oral fluid specimens. The review must
be completed before reporting the
results to the individual designated by
the licensee or other entity to assess
authorization or perform the suitability
and fitness determinations required
under § 26.619, or, if required, that are
described in subpart H of this part.
(2) Require their MRO to satisfy the
requirements in § 26.183 and, prior to
conducting any activities under this
part, attend and pass a medical- or
clinical-based training session to
improve his/her knowledge of MRO
duties and responsibilities, drug and
alcohol testing processes and
procedures, and evaluation of drug
testing results. This training session
must be conducted by a nationally
recognized MRO training and
certification organization that has been
assessed by the licensee’s or other
entity’s FFD program personnel to
include the technical elements an MRO
must implement under § 26.185. An

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MRO who performed the duties and
responsibilities in §§ 26.185 and 26.187
for at least 3 continuous years in the last
10 years prior to being hired or
contracted by the licensee or other
entity satisfies the requirements in this
paragraph.
(3) Require their MRO to attend a
medical- or clinical-based training
session on a triennial basis to improve
his/her knowledge of changes in drug
and alcohol testing processes and
procedures and evaluation of drug
testing results.
(4) Require their MRO to determine
whether a biological specimen is
positive, adulterated, substituted, dilute
or of questionable validity by
implementing the requirements in
§ 26.185 or the HHS Guidelines through
the licensee’s or other entity’s
procedures.
(i) If § 26.185 or the HHS Guidelines,
as used by the licensee or other entity
in its procedures, are insufficient to
make this determination, then guidance
issued by a State agency in the state in
which the NRC-licensed facility is
located, Federal agencies, or nationally
recognized MRO training and
certification organizations may be used
to inform an MRO determination.
(ii) An MRO need not review a
confirmed alcohol positive test result
determined by an evidentiary breath
testing device under paragraphs
(c)(3)(vi) and (vii) of this section.
(5) Require their MRO to determine
and approve the use of oral fluid or
urine as an alternative biological
specimen when the donor cannot
provide a specimen for testing. This
determination and the retest must be
documented and completed as soon as
reasonably practicable.
(6) Require the MRO to review all
specimens screened and tested
associated with a drug-related FFD
policy violation. This review includes
POCTA, split specimens, and all
specimens taken to resolve a discrepant
condition, such as a possible subversion
attempt, impairment without a known
cause, or a donor-requested or MROdirected re-test. To resolve a discrepant
condition, the MRO is authorized to test
a specimen for a biological marker,
adulterants, or additional drugs.
(n) Limitations of screening and
testing. Specimens collected under NRC
regulations may only be designated or
approved for screening and testing as
described in this part and may not be
used to conduct any other analysis or
test without the written permission of
the donor. Analyses, screens, and tests
that may not be conducted include, but
are not limited to, DNA testing,
serological typing, or any other medical

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or genetic test used for diagnostic or
specimen identification purposes. No
biological specimens may be passively
sampled and analyzed in a manner
different than described in this subpart.
(o) Specimen collectors. All onsite
specimen collections, except a
collection by a portal area screening
instrument in paragraph (j) of this
section, must be conducted by licenseeor other entity-designated and -trained
personnel.
§ 26.608

FFD program training.

(a) FFD program training. (1)
Individuals must be trained in the FFD
policy and procedure, including fatigue
management, and their FFD program
responsibilities. Individuals who collect
specimens for testing or screening must
also be trained in specimen collector
duties and responsibilities, including, at
a minimum, specimen collection,
custody and control, identification and
response to subversion attempts, and
privacy. For licensees and other entities
of commercial nuclear plants, the FFD
program training program must use a
systems approach to training as defined
in § 53.725 of this chapter and described
in § 53.830 of this chapter for those
individuals in § 26.4.
(2) FFD program training must
include training on the behavioral
observation program. The behavioral
observation program training must
include the detection of physiological
behaviors or conditions that may
indicate—
(i) Possible use, sale, or possession of
illegal drugs or illicit drugs, or
substance abuse on- or offsite;
(ii) Use or possession of alcohol onsite
or use while on duty offsite;
(iii) Impairment from fatigue or any
cause that, if left unattended, could
result in inattentiveness or human
errors; and
(iv) Any individual’s inability to
safely and competently perform
assigned duties and responsibilities or
act in a trustworthy and reliable manner
while having access to protected areas,
SNM, or sensitive information.
(3) Training must explain that an
individual’s FFD policy violation will—
(i) Subject the individual to an FFD
program-required sanction designed to
preclude recurrence of an FFD policy
violation;
(ii) Contribute to the licensee’s or
other entity’s assessment of whether the
individual can be trusted and relied
upon to safely and competently perform
the assigned duties and responsibilities
making the individual subject to this
subpart;
(iii) Be used to inform the licensee’s
or other entity’s insider mitigation and

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access authorization programs under
§§ 73.55, 73.56, 73.100 or 73.120 of this
chapter; and
(iv) Be used to inform other NRC
licensees and other entities subject to
part 26 when FFD program information
is requested to support authorization
determinations under subpart C of this
part or §§ 73.56 or 73.120 of this
chapter.
(b) Training and assessments.
Training and a trainee assessment must
be conducted before pre-access testing,
and refresher training and trainee
assessments must be conducted
periodically thereafter.
(c) Training program review. The
licensee or other entity must
periodically evaluate its FFD training
program and revise it as appropriate to
reflect industry experience as well as
applicable changes to the regulations in
this part, the HHS Guidelines, if used,
and specimen collection and testing
processes implemented by the licensee
or other entity.
§ 26.609

Behavioral observation.

(a) Licensees and other entities must
ensure that the individuals who are
subject to this subpart are subject to
behavioral observation and that
behavioral observation is performed by
all individuals subject to this subpart.
(b) Licensees and other entities must
require all individuals subject to the
FFD program to report to the licenseeor other entity-designated representative
any onsite or offsite behaviors or
activities by individuals subject to this
part that may constitute an
unreasonable risk to the safety or
security of the NRC-licensed facility or
SNM or may cause harm to others. This
reporting must include any information
relating to character or reputation of the
individual indicating that the individual
cannot be trusted or relied upon to
perform those duties and
responsibilities or maintain access to
NRC-licensed facilities, SNM, or
sensitive information that makes them
subject to part 26.
(c) Behavioral observation must be
performed visually, in-person, and,
when necessary, remotely by live video
and audible streaming and capture, to
observe the behavior of individuals in
the workforce subject to the
requirements in this subpart.
(d) Not withstanding paragraph (c) of
this section, for a reactor facility where
individual task loading does not allow
for the effective conduct of behavior
observation in addition to assigned
operational tasks, the licensee or other
entity must implement a live video and
audible streaming and capture system to
conduct behavioral observation of

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§ 26.613

persons licensed to operate under 10
CFR part 53 who manipulate the
controls of any commercial nuclear
plant licensed under 10 CFR part 53.
§ 26.610

Sanctions.

Licensees and other entities that
implement an FFD program under this
subpart must establish sanctions for
FFD policy violations that, at a
minimum, prohibit the individuals
specified in § 26.4 from being assigned
to perform or direct those duties and
responsibilities or maintaining
authorization making them subject to
this subpart. The severity of the
sanction must escalate with the number
of occurrences and severity of the FFD
policy violation. The sanction must be
long enough to act as a deterrent and,
if the individual is retained as a licensee
employee or contractor/vendor,
facilitate the individual to complete
counseling or treatment. The sanctions
must include a minimum 5-year denial
of access to the NRC-licensed facility for
any individual who is determined to
have been involved in the sale, use, or
possession of illegal drugs or the
consumption of alcohol within a
protected area of any facility licensed
under part 53 of this chapter or within
a transporter’s facility or vehicle used in
the conveyance of formula quantities of
strategic SNM while the individual is
subject to this subpart, and a permanent
denial of access to the NRC-licensed
facility for three FFD policy violations
or any subversion attempt of any drug
or alcohol test or screening process,
including subversion attempts at any
licensee or other entity subject to this
part.

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§ 26.611

Protection of information.

(a) Licensees and other entities that
collect personal information about an
individual for the purpose of complying
with this subpart must establish and
maintain a system of files and
procedures to prevent unauthorized
disclosure.
(b) Licensees and other entities must
obtain a signed consent that documents
the individual’s acceptance of being
subject to the FFD program and
authorizes the disclosure of the personal
information collected and maintained
under this subpart, except for
disclosures to the individuals and
entities specified in § 26.37(b)(1)
through (b)(6), (b)(8), and persons
deciding matters under review in
§ 26.613. This signed and dated consent
must be obtained before making the
individual subject to the FFD program.

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Appeals process.

Licensees and other entities that
implement an FFD program under this
subpart must establish and implement
procedures for the review of a
determination that an individual in
§ 26.4 has violated the FFD policy. The
procedure must provide for an objective
and impartial review of the facts related
to the determination that the individual
has violated the FFD policy and a
schedule for the completion of the
review.
§ 26.615

Audits.

(a) Licensees and other entities that
implement an FFD program under this
subpart must audit their programs at a
frequency that ensures the continuing
effectiveness of their FFD program, FFD
program elements that are provided by
C/Vs, and the FFD programs of C/Vs
that are accepted by the licensee or
other entity. Corrective actions must be
as soon as reasonably practicable to
resolve any problems identified in an
audit and preclude recurrence.
(b) The subject matter, scope, and
frequency of audits must be revised as
necessary to improve or maintain
program performance based on findings
resulting from licensee or other entity
implementation of its FFD PMRP in
§ 26.603(d).
(c) Licensees and other entities may
conduct joint audits or accept audits of
C/Vs so long as the audit addresses the
relevant services of the C/Vs.
(d) Licensees and other entities must
audit HHS-certified laboratories unless
the licensee’s or other entity’s panel of
drugs and drug metabolites to be tested
is equivalent to the panel by which the
laboratory is certified by HHS or is
subject to the standards and procedures
for drug testing and evaluation used by
the laboratory under the HHS
Guidelines. Licensees and other entities
must audit any hospital or other facility
licensed by the State (or Statedesignated entity) if used to conduct
specimen collections and perform
alcohol testing under this part on a
biennial basis to confirm that the facility
procedures are comparable to those
described in subpart E of this part, for
urine and oral fluid.
§ 26.617

Recordkeeping and reporting.

(a) Licensees and other entities that
implement FFD programs under this
subpart must ensure that records
pertaining to the administration of their
program, which may be stored and
archived electronically, are maintained
so that they are available for NRC
inspection purposes and for any legal
proceedings resulting from the
administration of the program. Records

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pertaining to the administration of the
FFD program and FFD performance data
required by § 26.717 must be retained
until license termination.
(b) Licensees and other entities must
make the following reports:
(1) Reports to the NRC Operations
Center by telephone within 24 hours
after the licensee or other entity
discovers any intentional act that casts
doubt on the integrity of the FFD
program and any programmatic failure,
degradation, or discovered vulnerability
of the FFD program that may permit
undetected drug or alcohol use or abuse
by individuals who are subject to this
subpart. These events must be reported
under this subpart, rather than under
the provisions of § 73.1200 of this
chapter; and
(2) Annual program performance
reports for the FFD program, including
the FFD program performance data
listed in § 26.717(b), as applicable.
Licensees and other entities must
submit FFD program performance data
(for January through December) to the
NRC annually, before March 1 of the
following year and must use unexpired
NRC-provided forms for the electronic
submission of FFD information to the
NRC.
(c) Licensees and other entities
subject to this subpart must describe in
sufficient detail to support an
authorization determination, an
individual’s FFD policy violation (while
protecting privacy information under
§ 26.611) and FFD program weakness to
NRC, licensees, and other entities
subject to this part when requested to
support authorization determinations
under subpart C of this part or § 73.120
of this chapter, as applicable, or to
support licensee or other entity
performance monitoring.
§ 26.619 Suitability and fitness
determinations.

Licensees and other entities that
implement FFD programs under this
subpart must develop, implement, and
maintain procedures for evaluating
whether to assign individuals to
perform or direct those duties and
responsibilities making them subject to
this subpart. A suitability or fitness
determination conducted for cause must
be performed face to face. A suitability
or fitness determination conducted for
cause may be performed remotely using
electronic communications only when
supported by someone who is present
in-person with the individual being
assessed, and that supporting person
must be trained in accordance with the
requirements of either §§ 26.29 or
26.608.
■ 94. Revise § 26.709 to read as follows:

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§ 26.709

Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules
Applicability.

(a) The requirements of this subpart
apply to the FFD programs of licensees
and other entities specified in § 26.3(a)
through (d), except for FFD programs
that are implemented under subpart K
of this part.
(b) The requirements in this subpart
apply to the FFD programs of licensees
and other entities specified in § 26.3(f)
that elect not to implement the
requirements in subpart M or elect to
implement the requirements in
§ 26.605(b).
§ 26.711

[Amended]

[Amended]

96. In § 26.825, in paragraph (b) add
remove the phrase ‘‘§§ 26.1, 26.3, 26.5,
26.7, 26.8, 26.9, 26.11, 26.51, 26.81,
26.121, 26.151, 26.181, 26.201, 26.823,
and 26.825’’ and add in its place the
phrase ‘‘§§ 26.1, 26.3, 26.5, 26.7, 26.8,
26.9, 26.11, 26.51, 26.81, 26.121, 26.151,
26.181, 26.201, 26.601, 26.823, and
26.825’’.

■

PART 30—RULES OF GENERAL
APPLICABILITY TO DOMESTIC
LICENSING OF BYPRODUCT
MATERIAL

Authority: Atomic Energy Act of 1954,
secs. 11, 81, 161, 181, 182, 183, 184, 186,
187, 223, 234, 274 (42 U.S.C. 2014, 2111,
2201, 2231, 2232, 2233, 2234, 2236, 2237,
2273, 2282, 2021); Energy Reorganization Act
of 1974, secs. 201, 202, 206, 211 (42 U.S.C.
5841, 5842, 5846, 5851); 44 U.S.C. 3504 note.

98. In § 30.4, revise the definition for
‘‘Utilization facility’’ to read as follows:

■

Definitions.

*

*
*
*
*
Utilization facility means a utilization
facility as defined in the regulations
contained in part 50 or part 53 of this
chapter;
■ 99. In § 30.50, revise paragraph (c)(3)
to read as follows:
Reporting requirements.

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*

*
*
*
*
(c) * * *
(3) The provisions of this section do
not apply to licensees subject to the
notification requirements in §§ 50.72 or
53.1630 of this chapter. They do apply
to those part 50 licensees possessing
material licensed under this part, who
are not subject to the notification
requirements in § 50.72 of this chapter.

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101. In § 40.60, revise paragraph (c)(3)
to read as follows:

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§ 40.60

Reporting requirements.

*

*
*
*
*
(c) * * *
(3) The provisions of this section do
not apply to licensees subject to the
notification requirements in §§ 50.72 or
53.1630 of this chapter. They do apply
to those part 50 licensees possessing
material licensed under this part who
are not subject to the notification
requirements in § 50.72 of this chapter.
PART 50—DOMESTIC LICENSING OF
PRODUCTION AND UTILIZATION
FACILITIES
102. The authority citation for part 50
continues to read as follows:

97. The authority citation for part 30
continues to read as follows:

§ 30.50

Authority: Atomic Energy Act of 1954,
secs. 62, 63, 64, 65, 69, 81, 83, 84, 122, 161,
181, 182, 183, 184, 186, 187, 193, 223, 234,
274, 275 (42 U.S.C. 2092, 2093, 2094, 2095,
2099, 2111, 2113, 2114, 2152, 2201, 2231,
2232, 2233, 2234, 2236, 2237, 2243, 2273,
2282, 2021, 2022); Energy Reorganization Act
of 1974, secs. 201, 202, 206, 211 (42 U.S.C.
5841, 5842, 5846, 5851); Uranium Mill
Tailings Radiation Control Act of 1978, sec.
104 (42 U.S.C. 7914); 44 U.S.C. 3504 note.

■

■

§ 30.4

100. The authority citation for part 40
continues to read as follows:

■

■

95. In § 26.711, in paragraphs (c) and
(d), remove the phrase ‘‘(c) and (d),’’ and
add in its place the phrase ‘‘(c), (d), and
(f),’’.

■

§ 26.825

PART 40—DOMESTIC LICENSING OF
SOURCE MATERIAL

Authority: Atomic Energy Act of 1954,
secs. 11, 101, 102, 103, 104, 105, 108, 122,
147, 149, 161, 181, 182, 183, 184, 185, 186,
187, 189, 223, 234 (42 U.S.C. 2014, 2131,
2132, 2133, 2134, 2135, 2138, 2152, 2167,
2169, 2201, 2231, 2232, 2233, 2234, 2235,
2236, 2237, 2239, 2273, 2282); Energy
Reorganization Act of 1974, secs. 201, 202,
206, 211 (42 U.S.C. 5841, 5842, 5846, 5851);
Nuclear Waste Policy Act of 1982, sec.
306(42 U.S.C. 10226); National
Environmental Policy Act of 1969 (42 U.S.C.
4332); 44 U.S.C. 3504 note; Sec. 109, Pub. L.
96–295, 94 Stat. 783.

103. In § 50.47, revise paragraphs
(a)(1) and (e) to read as follows:

■

§ 50.47

Emergency plans.

(a)(1)(i) Except as provided in
paragraph (d) of this section, no initial
operating license for a nuclear power
reactor will be issued under this part or
under part 53 of this chapter unless a
finding is made by the NRC that there
is reasonable assurance that adequate
protective measures can and will be
taken in the event of a radiological
emergency. No finding under this
section is necessary for issuance of a
renewed nuclear power reactor
operating license.
(ii) No initial combined license under
parts 52 or 53 of this chapter will be
issued unless a finding is made by the

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NRC that there is reasonable assurance
that adequate protective measures can
and will be taken in the event of a
radiological emergency. No finding
under this section is necessary for
issuance of a renewed combined
license.
(iii) If an application for an early site
permit under subpart A of part 52 of this
chapter includes complete and
integrated emergency plans under
§ 52.17(b)(2)(ii) of this chapter or an
application for an early site permit
under subpart H of part 53 of this
chapter includes complete and
integrated emergency plans under
§ 53.1146(b)(2)(ii) of this chapter, no
early site permit will be issued unless
a finding is made by the NRC that the
emergency plans provide reasonable
assurance that adequate protective
measures can and will be taken in the
event of a radiological emergency.
(iv) If an application for an early site
permit proposes major features of the
emergency plans under §§ 52.17(b)(2)(i)
or 53.1146(b)(2)(i) of this chapter, no
early site permit will be issued unless
a finding is made by the NRC that the
major features are acceptable in
accordance with the applicable
standards of either § 50.47 and appendix
E to this part or the applicable
requirements of § 50.160, within the
scope of emergency preparedness
matters addressed in the major features.
*
*
*
*
*
(e) Notwithstanding the requirements
of paragraph (b) of this section and the
provisions of § 52.103 or § 53.1452 of
this chapter, a holder of a combined
license under part 52 or part 53 of this
chapter, as applicable, that is complying
with the requirements of § 50.47(b) and
appendix E to this part may not load
fuel or operate except as provided in
accordance with appendix E to this part
and § 50.54(gg), and a holder of a
combined license under part 52 or part
53 of this chapter that is complying with
the requirements of § 50.160 may not
load fuel or operate except as provided
in accordance with § 50.160(c)(2) and
§ 50.54(gg).
*
*
*
*
*
■ 104. In § 50.54, revise paragraphs
(q)(2), (q)(4), and (gg)(1) introductory
text to read as follows:
§ 50.54

Conditions of licenses.

*

*
*
*
*
(q) * * *
(2)(i) Except as provided in paragraph
(q)(2)(ii) of this section, a holder of a
license under this part, or a combined
license under parts 52 or 53 of this
chapter after the Commission makes the
finding under §§ 52.103(g) or 53.1452(g)

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of this chapter, as applicable, shall
follow and maintain the effectiveness of
an emergency plan that meets the
requirements in appendix E to this part
and, for nuclear power reactor licensees,
the planning standards of § 50.47(b).
(ii) A holder of a license under this
part for a non-power production or
utilization facility, a holder of a license
under this part or part 53 of this chapter
for a small modular reactor or a nonlight-water reactor, or a holder of a
combined license under parts 52 or 53
of this chapter after the Commission
makes the finding under §§ 52.103(g) or
53.1452(g) of this chapter, as applicable,
for a small modular reactor or a nonlight-water reactor, shall follow and
maintain the effectiveness of either an
emergency plan that meets the
requirements in § 50.160 or an
emergency plan that meets the
requirements in appendix E to this part
and, for nuclear power reactor licensees,
the planning standards of § 50.47(b).
*
*
*
*
*
(4) The changes to a licensee’s
emergency plan that reduce the
effectiveness of the plan as defined in
paragraph (q)(1)(iv) of this section may
not be implemented without prior
approval by the NRC. A licensee
desiring to make such a change shall
submit an application for an
amendment to its license. In addition to
the filing requirements of §§ 50.90 and
50.91 or §§ 53.1510 and 53.1515 of this
chapter, as applicable, the request must
include all emergency plan pages
affected by that change and must be
accompanied by a forwarding letter
identifying the change, the reason for
the change, and the basis for concluding
that the licensee’s emergency plan, as
revised, will continue to meet either the
requirements in § 50.160 or the
requirements in appendix E to this part
and, for nuclear power reactor licensees,
the planning standards of § 50.47(b).
*
*
*
*
*
(gg)(1) Notwithstanding §§ 52.103 or
53.1452 of this chapter, if following the
conduct of the exercise required by
paragraph IV.f.2.a of appendix E to this
part or § 50.160(c)(2), as applicable,
FEMA identifies one or more
deficiencies in the state of offsite
emergency preparedness, the holder of a
combined license under 10 CFR part 52
or under 10 CFR part 53, as applicable,
may operate at up to 5 percent of rated
thermal power only if the Commission
finds that the state of onsite emergency
preparedness provides reasonable
assurance that adequate protective
measures can and will be taken in the
event of a radiological emergency. The
NRC will base this finding on its

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assessment of the applicant’s onsite
emergency plans against the pertinent
standards in either § 50.47 and
appendix E to this part, or § 50.160, as
applicable. Review of the applicant’s
emergency plans will include the
following standards with offsite aspects:
*
*
*
*
*
■ 105. In § 50.160, revise paragraphs
(b)(3) and (c)(2) to read as follows:
§ 50.160 Emergency preparedness for
small modular reactors, non-light-water
reactors, and non-power production or
utilization facilities.

*

*
*
*
*
(b) * * *
(3) Emergency planning zone. For an
applicant whose analysis required by
§ 50.33(g)(2) or § 53.1109(g)(2) of this
chapter meets the criteria in
§ 50.33(g)(2)(i) or § 53.1109(g)(2)(i) of
this chapter, as applicable, determine
and describe the boundary and physical
characteristics of the EPZ in the
emergency plan.
*
*
*
*
*
(c) * * *
(2) A holder of a combined license
issued under parts 52 or 53 of this
chapter before the Commission has
made the finding under §§ 52.103(g) or
53.1452(g) of this chapter, as applicable,
must establish, implement, and
maintain an emergency preparedness
program that meets the requirements of
paragraph (b) of this section, as
described in the approved emergency
plan and license, and conduct an initial
exercise to demonstrate this compliance
within 2 years before the scheduled date
for initial loading of fuel (or, for a fueled
manufactured reactor, within 2 years
before the scheduled date for initiating
the physical removal of any one of the
independent physical mechanisms to
prevent criticality required under
§ 53.620(d)(1) of this chapter).
■ 106. In appendix B to part 50, revise
the first paragraph in the Introduction
section, the first paragraph of section III,
Design Control, and section IV,
Procurement Document Control, to read
as follows:
Appendix B to Part 50—Quality
Assurance Criteria for Nuclear Power
Plants and Fuel Reprocessing Plants
Introduction. Every applicant for a
construction permit is required by the
provisions of § 50.34 or § 53.1309 of this
chapter to include in its preliminary safety
analysis report a description of the quality
assurance program to be applied to the
design, fabrication, construction, and testing
of the structures, systems, and components of
the facility. Every applicant for an operating
license is required by the provisions of
§ 50.34 or § 53.1369 of this chapter to
include, in its final safety analysis report,

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87039

information pertaining to the managerial and
administrative controls to be used to assure
safe operation. Every applicant for a
combined license is required by the
provisions of §§ 52.79 or 53.1416 of this
chapter to include in its final safety analysis
report a description of the quality assurance
applied to the design, and to be applied to
the fabrication, construction, and testing of
the structures, systems, and components of
the facility and to the managerial and
administrative controls to be used to assure
safe operation. For applications submitted
after September 27, 2007, every applicant for
an early site permit is required by the
provisions of §§ 52.17 or 53.1146 of this
chapter to include in its site safety analysis
report a description of the quality assurance
program applied to site activities related to
the design, fabrication, construction, and
testing of the structures, systems, and
components of a facility or facilities that may
be constructed on the site. Every applicant
for a design approval is required by the
provisions of §§ 52.137 or 53.1209 of this
chapter to include in its final safety analysis
report a description of the quality assurance
program applied to the design of the
structures, systems, and components of the
facility. Every applicant for a design
certification is required by the provisions of
§§ 52.47 or 53.1239 of this chapter to include
in its final safety analysis report a description
of the quality assurance program applied to
the design of the structures, systems, and
components of the facility. Every applicant
for a manufacturing license is required by the
provisions of §§ 52.157 or 53.1279 of this
chapter to include in its final safety analysis
report a description of the quality assurance
program applied to the design, and to be
applied to the manufacture of, the structures,
systems, and components of the reactor.
Nuclear power plants and fuel reprocessing
plants include structures, systems, and
components that prevent or mitigate the
consequences of postulated accidents that
could cause undue risk to the health and
safety of the public. This appendix
establishes quality assurance requirements
for the design, manufacture, construction,
and operation of those structures, systems,
and components. The pertinent requirements
of this appendix apply to all activities
affecting the safety-related functions of those
structures, systems, and components; these
activities include designing, purchasing,
fabricating, handling, shipping, storing,
cleaning, erecting, installing, inspecting,
testing, operating, maintaining, repairing,
refueling, and modifying.

*

*

*

*

*

III. Design Control
Measures shall be established to assure that
applicable regulatory requirements and the
design bases, as defined in § 50.2 and as
specified in the license application, or the
functional design criteria, as defined in
§ 53.020 of this chapter and as specified in
the license application, for those structures,
systems, and components to which this
appendix applies are correctly translated into
specifications, drawings, procedures, and
instructions. These measures shall include
provisions to assure that appropriate quality

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standards are specified and included in
design documents and that deviations from
such standards are controlled. Measures shall
also be established for the selection and
review for suitability of application of
materials, parts, equipment, and processes
that are essential to the safety-related
functions of the structures, systems and
components.

*

*

*

*

*

IV. Procurement Document Control
Measures shall be established to assure that
applicable regulatory requirements, design
bases or functional design criteria, and other
requirements which are necessary to assure
adequate quality are suitably included or
referenced in the documents for procurement
of material, equipment, and services, whether
purchased by the applicant or by its
contractors or subcontractors. To the extent
necessary, procurement documents shall
require contractors or subcontractors to
provide a quality assurance program
consistent with the pertinent provisions of
this appendix.

*

*

*

*

*

PART 51—ENVIRONMENTAL
PROTECTION REGULATIONS FOR
DOMESTIC LICENSING AND RELATED
REGULATORY FUNCTIONS
107. The authority citation for part 51
continues to read as follows:

■

Authority: Atomic Energy Act of 1954,
secs. 161, 193 (42 U.S.C. 2201, 2243); Energy
Reorganization Act of 1974, secs. 201, 202
(42 U.S.C. 5841, 5842); National
Environmental Policy Act of 1969 (42 U.S.C.
4332, 4334, 4335); Nuclear Waste Policy Act
of 1982, secs. 144(f), 121, 135, 141, 148 (42
U.S.C. 10134(f), 10141, 10155, 10161, 10168);
44 U.S.C. 3504 note.

108. In § 51.20, revise paragraphs
(b)(1) and (2) to read as follows:

■

§ 51.20 Criteria for and identification of
licensing and regulatory actions requiring
environmental impact statements.

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*

*
*
*
*
(b) * * *
(1) Issuance of a limited work
authorization or a permit to construct a
nuclear power reactor, testing facility, or
fuel reprocessing plant under part 50 of
this chapter, issuance of an early site
permit under part 52 of this chapter, or
issuance of a limited work
authorization, construction permit, or
early site permit under part 53 of this
chapter.
(2) Issuance or renewal of a full power
or design capacity license to operate a
nuclear power reactor, testing facility, or
fuel reprocessing plant under parts 50 or
53 of this chapter, or a combined license
under parts 52 or 53 of this chapter.
*
*
*
*
*
■ 109. In § 51.22, revise paragraphs
(c)(3) introductory text, (c)(9)
introductory text, (c)(12) introductory

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text, (c)(17), (c)(22) and (23) to read as
follows:
§ 51.22 Criterion for categorical exclusion;
identification of licensing and regulatory
actions eligible for categorical exclusion or
otherwise not requiring environmental
review.

*

*
*
*
*
(c) * * *
(3) Amendments to parts 20, 30, 31,
32, 33, 34, 35, 37, 39, 40, 50, 51, 52, 53,
54, 60, 61, 63, 70, 71, 72, 73, 74, 81, and
100 of this chapter which relate to—
*
*
*
*
*
(9) Issuance of an amendment to a
permit or license for a reactor under part
50, part 52, or part 53 of this chapter
that changes a requirement or issuance
of an exemption from a requirement,
with respect to installation or use of a
facility component located within the
restricted area, as defined in part 20 of
this chapter; or the issuance of an
amendment to a permit or license for a
reactor under part 50, part 52, or part 53
of this chapter that changes an
inspection or a surveillance
requirement; provided that:
*
*
*
*
*
(12) Issuance of an amendment to a
license under parts 50, 52, 53, 60, 61,
63, 70, 72, or 75 of this chapter relating
solely to safeguards matters (i.e.,
protection against sabotage or loss or
diversion of special nuclear material) or
issuance of an approval of a safeguards
plan submitted under parts 50, 52, 53,
70, 72, and 73 of this chapter, provided
that the amendment or approval does
not involve any significant construction
impacts. These amendments and
approvals are confined to—
*
*
*
*
*
(17) Issuance of an amendment to a
permit or license under part 30, part 40,
part 50, part 52, part 53, or part 70 of
this chapter which deletes any limiting
condition of operation or monitoring
requirement based on or applicable to
any matter subject to the provisions of
the Federal Water Pollution Control Act.
*
*
*
*
*
(22) Issuance of a standard design
approval under part 52 or part 53 of this
chapter.
(23) The Commission finding for a
combined license under § 52.103(g) or
§ 53.1452(g) of this chapter.
*
*
*
*
*
§ 51.26

[Amended]

110. In § 51.26, in paragraph (d)
remove the phrase ‘‘under part 52’’ and
add in its place the phrase ‘‘under 10
CFR parts 52 or 53,’’.
■ 111. In § 51.30, revise paragraph (a)
introductory text and paragraphs (d) and
(e) to read as follows:
■

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§ 51.30

Environmental assessment.

(a) An environmental assessment for
proposed actions, other than those for a
standard design certification under 10
CFR parts 52 or 53, or a manufacturing
license under 10 CFR parts 52 or 53,
shall identify the proposed action and
include:
*
*
*
*
*
(d) An environmental assessment for
a standard design certification under
subpart B of part 52 of this chapter, or
under subpart H of part 53 of this
chapter must identify the proposed
action and will be limited to the
consideration of the costs and benefits
of severe accident mitigation design
alternatives and the bases for not
incorporating severe accident mitigation
design alternatives in the design
certification. An environmental
assessment for an amendment to a
design certification will be limited to
the consideration of whether the design
change which is the subject of the
proposed amendment renders a severe
accident mitigation design alternative
previously rejected in the earlier
environmental assessment to become
cost beneficial, or results in the
identification of new severe accident
mitigation design alternatives, in which
case the costs and benefits of new severe
accident mitigation design alternatives
and the bases for not incorporating new
severe accident mitigation design
alternatives in the design certification
must be addressed.
(e) An environmental assessment for a
manufacturing license under subpart F
of part 52 of this chapter or under
subpart H of part 53 of this chapter must
identify the proposed action and will be
limited to the consideration of the costs
and benefits of severe accident
mitigation design alternatives and the
bases for not incorporating severe
accident mitigation design alternatives
in the manufacturing license. An
environmental assessment for an
amendment to a manufacturing license
will be limited to consideration of
whether the design change which is the
subject of the proposed amendment
either renders a severe accident
mitigation design alternative previously
rejected in an environmental assessment
to become cost beneficial, or results in
the identification of new severe accident
mitigation design alternatives, in which
case the costs and benefits of new severe
accident mitigation design alternatives
and the bases for not incorporating new
severe accident mitigation design
alternatives in the manufacturing
license must be addressed. In either
case, the environmental assessment will
not address the environmental impacts

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associated with manufacturing the
reactor under the manufacturing license.
§ 51.31

[Amended]

112. In § 51.31, in paragraph (a)
remove the phrase ‘‘under part 52’’ and
add in its place the phrase ‘‘under parts
52 or 53’’.

■

§ 51.32

[Amended]

113. In § 51.32, in paragraphs (b)(1)
and (3) remove the phrase ‘‘of part 52 of
this chapter’’ and add in its place the
phrase ‘‘of part 52 of this chapter or
subpart H of part 53 of this chapter’’.

■

§ 51.49

[Amended]

114. In § 51.49, in paragraph (c)
introductory text, remove the phrase ‘‘of
part 52 of this chapter’’ and add in its
place the phrase ‘‘of part 52 of this
chapter or under subpart H of part 53 of
this chapter’’.

■

§ 51.50

[Amended]

115. In § 51.50, wherever it appears,
remove the phrase ‘‘in accordance with
§ 50.36b of this chapter’’ and add in its
place the phrase ‘‘in accordance with
§§ 50.36b or 53.1112 of this chapter’’.

■

§ 51.53

[Amended]

116. In § 51.53, in paragraph (d)
remove the phrase ‘‘under § 50.82 of this
chapter’’ and add in its place the phrase
‘‘under §§ 50.82 or 53.1080 of this
chapter’’.

■

§ 51.54

[Amended]

117. In § 51.54, in paragraph (a),
remove the phrase ‘‘of part 52 of this
chapter’’ and add in its place the phrase
‘‘of part 52 of this chapter or under
subpart H of part 53 of this chapter’’.

■

§ 51.55

[Amended]

118. In § 51.55, in paragraph (a)
remove the phrase ‘‘of part 52 of this
chapter’’ and add in its place the phrase
‘‘of part 52 of this chapter or under
subpart H of part 53 of this chapter’’.
■ 119. In § 51.58, revise paragraph (b) to
read as follows:
■

§ 51.58 Environmental report—number of
copies; distribution.

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*

*
*
*
*
(b) Each applicant for a license to
manufacture a nuclear power reactor, or
for an amendment to a license to
manufacture, seeking approval of the
final design of the nuclear power reactor
under subpart F of part 52 of this
chapter or under subpart H of part 53 of
this chapter, shall submit to the
Commission an environmental report or
any supplement to an environmental
report in the manner specified in §§ 52.3
or 53.040 of this chapter. The applicant
shall maintain the capability to generate

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additional copies of the environmental
report or any supplement to the
environmental report for subsequent
distribution to parties and Boards in the
NRC proceeding; Federal, State, and
local officials; and any affected Indian
Tribes, in accordance with written
instructions issued by the Director,
Office of Nuclear Reactor Regulation.
■ 120. In § 51.77, revise paragraph (a)
introductory text to read as follows:
§ 51.77 Distribution of draft environmental
impact statement.

(a) In addition to the distribution
authorized by § 51.74, a copy of a draft
environmental statement for a licensing
action for a production or utilization
facility, except an action authorizing
issuance, amendment, or renewal of a
license to manufacture a nuclear power
reactor pursuant to 10 CFR part 52,
subpart F or 10 CFR part 53, subparts H
or I will also be distributed to:
*
*
*
*
*
§ 51.92

[Amended]

121. In § 51.92, in paragraph (b),
wherever it may appear, remove the
phrase ‘‘10 CFR part 52’’ and add in its
place the phrase ‘‘10 CFR parts 52 or
53’’.

■

§ 51.95

[Amended]

122. In § 51.95, in paragraph (c)
introductory text remove the phrase
‘‘under 10 CFR parts 52 or 54’’ and add
in its place the phrase ‘‘under 10 CFR
parts 52, 53, or 54’’.
■ 123. In § 51.101, revise paragraph
(a)(2) to read as follows:
■

§ 51.101

Limitations on actions.

(a) * * *
(2) Any action concerning the
proposal taken by an applicant which
would—
(i) Have an adverse environmental
impact, or
(ii) Limit the choice of reasonable
alternatives that may be grounds for
denial of the license. In the case of an
application covered by §§ 30.32(f),
40.31(f), 50.10(c), 53.1130, 70.21(f), or
72.16 and 72.34 of this chapter, the
provisions of this paragraph will be
applied in accordance with
§ 30.33(a)(5), 40.32(e), 50.10(c), 53.1130,
70.23(a)(7), or 72.40(b) of this chapter,
as appropriate.
*
*
*
*
*
§ 51.103

[Amended]

124. In § 51.103, in paragraph (a)(6)
remove the phrase ‘‘under 10 CFR
50.10’’ and add in its place the phrase
‘‘under §§ 50.10 or 53.1130 of this
chapter’’.

■

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125. In § 51.105, revise paragraph
(c)(1) introductory text to read as
follows:

■

§ 51.105 Public hearings in proceedings
for issuance of construction permits or
early site permits; limited work
authorizations.

*

*
*
*
*
(c)(1) In addition to complying with
the applicable provisions of § 51.104, in
any proceeding for the issuance of a
construction permit for a nuclear power
plant or an early site permit under parts
52 or 53 of this chapter, where the
applicant requests a limited work
authorization under §§ 50.10(d) or
53.1130 of this chapter, the presiding
officer will—
*
*
*
*
*
■ 126. In § 51.107, revise paragraphs (a)
introductory text, (b) introductory text,
and (d)(1) introductory text to read as
follows:
§ 51.107 Public hearings in proceedings
for issuance of combined licenses; limited
work authorizations.

(a) In addition to complying with the
applicable requirements of § 51.104, in
a proceeding for the issuance of a
combined license for a nuclear power
reactor under parts 52 or 53 of this
chapter, the presiding officer will:
*
*
*
*
*
(b) If a combined license application
references an early site permit, then the
presiding officer in the combined
license hearing must not admit any
contention proffered by any party on
environmental issues that have been
accorded finality under §§ 52.39 or
53.1188 of this chapter, unless the
contention:
*
*
*
*
*
(d)(1) In any proceeding for the
issuance of a combined license where
the applicant requests a limited work
authorization under §§ 50.10(d) or
§ 53.1130(a) of this chapter, the
presiding officer, in addition to
complying with any applicable
provision of § 51.104, will:
*
*
*
*
*
■ 127. Revise § 51.108 to read as
follows:
§ 51.108 Public hearings on Commission
findings that inspections, tests, analyses,
and acceptance criteria of combined
licenses are met.

In any public hearing requested under
§§ 52.103(b) or 53.1452(b) of this
chapter, the Commission will not admit
any contentions on environmental
issues, the adequacy of the
environmental impact statement for the
combined license issued under subpart
C of part 52 of this chapter or under

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subpart H of part 53 of this chapter, or
the adequacy of any other
environmental impact statement or
environmental assessment referenced in
the combined license application. The
Commission will not make any
environmental findings in connection
with the finding under § 52.103(g) or
§ 53.1452(g) of this chapter.
■ 128. Add part 53, consisting of
§§ 53.000 through 53.9010, to read as
follows:
PART 53—RISK-INFORMED,
TECHNOLOGY-INCLUSIVE
REGULATORY FRAMEWORK FOR
COMMERCIAL NUCLEAR PLANTS
Sec.
53.000

Purpose.

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Subpart B—Technology-Inclusive Safety
Requirements
53.210 Safety criteria for design-basis
accidents.
53.220 Safety criteria for licensing-basis
events other than design-basis accidents.
53.230 Safety functions.
53.240 Licensing-basis events.
53.250 Defense in depth.
53.260 Normal operations.
53.270 Protection of plant workers.
Subpart C—Design and Analysis
Requirements
53.400 Design features for licensing-basis
events.
53.410 Functional design criteria for
design-basis accidents.
53.415 Protection against external hazards.
53.420 Functional design criteria for
licensing-basis events other than designbasis accidents.
53.425 Design features and functional
design criteria for normal operations.
53.430 Design features and functional
design criteria for protection of plant
workers.
53.440 Design requirements.
53.450 Analysis requirements.
53.460 Safety categorization and treatments.
53.470 Maintaining analytical safety
margins used to justify operational
flexibilities.
53.480 Earthquake engineering.

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53.500
53.510
53.520
53.530
53.540

General siting and siting assessment.
External hazards.
Site characteristics.
Population-related considerations.
Siting interfaces.

Subpart E—Construction and
Manufacturing Requirements
53.600 Construction and manufacturing—
scope and purpose.
53.605 Reporting of defects and
noncompliance.
53.610 Construction.
53.620 Manufacturing.
Subpart F—Requirements for Operation

Subpart A—General Provisions
53.015 Scope.
53.020 Definitions.
53.030 Reserved.
53.040 Written communications.
53.050 Deliberate misconduct.
53.060 Employee protection.
53.070 Completeness and accuracy of
information.
53.080 Specific exemptions.
53.090 Standards for review.
53.100 Jurisdictional limits.
53.110 Attacks and destructive acts.
53.115 Rights related to special nuclear
material.
53.117 License suspension and rights of
recapture.
53.120 Information collection requirements:
OMB approval.

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Subpart D—Siting Requirements

53.700 Operational objectives.
53.710 Maintaining capabilities and
availability of structures, systems, and
components.
53.715 Maintenance, repair, and inspection
programs.
53.720 Response to seismic events.
53.725 General staffing, training, personnel
qualifications, and human factors
requirements.
53.726 Communications.
53.728 Completeness and accuracy of
information.
53.730 Defining, fulfilling, and maintaining
the role of personnel in ensuring safe
operations.
53.735 General exemptions.
53.740 Facility licensee requirements—
General.
53.745 Operator license requirements.
53.760 Operator licensing.
53.765 Medical requirements.
53.770 Incapacitation because of disability
or illness.
53.775 Applications for operators and
senior operators.
53.780 Training, examination, and
proficiency program.
53.785 Conditions of operator and senior
operator licenses.
53.790 Issuance, modification, and
revocation of operator and senior
operator licenses.
53.795 Expiration and renewal of operator
and senior operator licenses.
53.800 Facility licensees for self-reliantmitigation facilities.
53.805 Facility licensee requirements
related to generally licensed reactor
operators.
53.810 Generally licensed reactor operators.
53.815 Generally licensed reactor operator
training, examination, and proficiency
programs.
53.820 Cessation of individual
applicability.
53.830 Training and qualification of
commercial nuclear plant personnel.
53.845 Programs.
53.850 Radiation protection.
53.855 Emergency preparedness.
53.860 Security programs.
53.865 Quality assurance.
53.870 Integrity assessment programs.
53.875 Fire protection.
53.880 Inservice inspection and inservice
testing.
53.910 Procedures and guidelines.

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Subpart G—Decommissioning
Requirements
53.1000 Scope and purpose.
53.1010 Financial assurance for
decommissioning.
53.1020 Cost estimates for
decommissioning.
53.1030 Annual adjustments to cost
estimates for decommissioning.
53.1040 Methods for providing financial
assurance for decommissioning.
53.1045 Limitations on the use of
decommissioning trust funds.
53.1050 NRC oversight.
53.1060 Reporting and recordkeeping
requirements.
53.1070 Termination of license.
53.1075 Program requirements during
decommissioning.
53.1080 Release of part of a commercial
nuclear plant or site for unrestricted use.
Subpart H—Licenses, Certifications, and
Approvals
53.1100 Filing of application for licenses,
certifications, or approvals; oath or
affirmation.
53.1101 Requirement for license.
53.1103 Combining applications and
licenses.
53.1106 Elimination of repetition.
53.1109 Contents of applications; general
information.
53.1112 Environmental conditions.
53.1115 Agreement limiting access to
classified information.
53.1118 Ineligibility of certain applicants.
53.1120 Exceptions and exemptions from
licensing requirements.
53.1121 Public inspection of applications.
53.1124 Relationship between sections.
53.1130 Limited work authorizations.
53.1140 Early site permits.
53.1143 Filing of applications.
53.1144 Contents of applications for early
site permits; general information.
53.1146 Contents of applications for early
site permits; technical information.
53.1149 Review of applications.
53.1155 Referral to the Advisory Committee
on Reactor Safeguards.
53.1158 Issuance of early site permit.
53.1161 Extent of activities permitted.
53.1164 Duration of permit.
53.1167 Limited work authorization after
issuance of early site permit.
53.1170 Transfer of early site permit.
53.1173 Application for renewal.
53.1176 Criteria for renewal.
53.1179 Duration of renewal.
53.1182 Use of site for other purposes.
53.1188 Finality of early site permit
determinations.
53.1200 Standard design approvals.
53.1203 Filing of applications.
53.1206 Contents of applications for
standard design approvals; general
information.
53.1209 Contents of applications for
standard design approvals; technical
information.
53.1210 Contents of applications for
standard design approvals; other
application content.
53.1212 Standards for review of
applications.

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53.1215 Referral to the Advisory Committee
on Reactor Safeguards.
53.1218 Staff approval of design.
53.1221 Finality of standard design
approvals; information requests.
53.1230 Standard design certifications.
53.1233 Filing of applications.
53.1236 Contents of applications for
standard design certifications; general
information.
53.1239 Contents of applications for
standard design certifications; technical
information.
53.1241 Contents of applications for
standard design certifications; other
application content.
53.1242 Review of applications.
53.1245 Referral to the Advisory Committee
on Reactor Safeguards.
53.1248 Issuance of standard design
certification.
53.1251 Duration of certification.
53.1254 Application for renewal.
53.1257 Criteria for renewal.
53.1260 Duration of renewal.
53.1263 Finality of standard design
certifications.
53.1270 Manufacturing licenses.
53.1273 Filing of applications.
53.1276 Contents of applications for
manufacturing licenses; general
information.
53.1279 Contents of applications for
manufacturing licenses; technical
information.
53.1282 Contents of applications for
manufacturing licenses; other
application content.
53.1285 Review of applications.
53.1286 Referral to the Advisory Committee
on Reactor Safeguards.
53.1287 Issuance of manufacturing licenses.
53.1288 Finality of manufacturing licenses.
53.1291 Duration of manufacturing
licenses.
53.1293 Transfer of manufacturing licenses.
53.1295 Renewal of manufacturing licenses.
53.1300 Construction permits.
53.1306 Contents of applications for
construction permits; general
information.
53.1309 Contents of applications for
construction permits; technical
information.
53.1312 Contents of applications for
construction permits; other application
content.
53.1315 Review of applications.
53.1318 Finality of referenced NRC
approvals, permits, and certifications.
53.1324 Referral to the Advisory Committee
on Reactor Safeguards.
53.1327 Authorization to conduct limited
work authorization activities.
53.1330 Exemptions, departures, and
variances.
53.1333 Issuance of construction permits.
53.1336 Finality of construction permits.
53.1342 Duration of construction permits.
53.1345 Transfer of construction permits.
53.1348 Termination of construction
permits.
53.1360 Operating licenses.
53.1366 Contents of applications for
operating licenses; general information.

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53.1369 Contents of applications for
operating licenses; technical
information.
53.1372 Contents of applications for
operating licenses; other application
content.
53.1375 Review of applications.
53.1381 Referral to the Advisory Committee
on Reactor Safeguards.
53.1384 Exemptions, departures, and
variances.
53.1387 Issuance of operating licenses.
53.1390 Backfitting of operating licenses.
53.1396 Duration of operating licenses.
53.1399 Transfer of an operating license.
53.1402 Application for renewal.
53.1405 Continuation of an operating
license.
53.1410 Combined licenses.
53.1413 Contents of applications for
combined licenses; general information.
53.1416 Contents of applications for
combined licenses; technical
information.
53.1419 Contents of applications for
combined licenses; other application
content.
53.1422 Review of applications.
53.1425 Finality of referenced NRC
approvals.
53.1431 Referral to the Advisory Committee
on Reactor Safeguards.
53.1434 Authorization to conduct limited
work authorization activities.
53.1437 Exemptions, departures, and
variances.
53.1440 Issuance of combined licenses.
53.1443 Finality of combined licenses.
53.1449 Inspection during construction.
53.1452 Operation under a combined
license.
53.1455 Duration of combined license.
53.1456 Transfer of a combined license.
53.1458 Application for renewal.
53.1461 Continuation of combined license.
53.1470 Standardization of commercial
nuclear plant designs: licenses to
construct and operate nuclear power
reactors of identical design at multiple
sites.
Subpart I—Maintaining and Revising
Licensing-Basis Information
53.1500 Licensing-basis information.
53.1502 Specific terms and conditions of
licenses.
53.1505 Changes to licensing-basis
information requiring prior NRC
approval.
53.1510 Application for amendment of
license.
53.1515 Public notices; State consultation.
53.1520 Issuance of amendment.
53.1525 Revising certification information
within a design certification rule.
53.1530 Revising design information within
a manufacturing license.
53.1535 Amendments during construction.
53.1540 Updating licensing-basis
information and determining the need
for NRC approval.
53.1545 Updating Final Safety Analysis
Reports.
53.1550 Evaluating changes to facility as
described in Final Safety Analysis
Reports.

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53.1560 Updating program documents
included in licensing-basis information.
53.1565 Evaluating changes to programs
included in licensing-basis information.
53.1570 Transfer of licenses.
53.1575 Termination of licenses.
53.1580 Information requests.
53.1585 Revocation, suspension,
modification of licenses and approvals
for cause.
53.1590 Backfitting.
53.1595 Renewal.
Subpart J—Reporting and Other
Administrative Requirements
53.1600 General information.
53.1610 Unfettered access for inspections.
53.1620 Maintenance of records, making of
reports.
53.1630 Immediate notification
requirements for operating commercial
nuclear plants.
53.1640 Licensee event report system.
53.1645 Reports of radiation exposure to
members of the public.
53.1650 Facility information and
verification.
53.1660 Financial requirements.
53.1670 Financial qualifications.
53.1680 Annual financial reports.
53.1690 Licensee’s change of status;
financial qualifications.
53.1700 Creditor regulations.
53.1710 Financial protection.
53.1720 Insurance required to stabilize and
decontaminate plant following an
accident.
53.1730 Financial protection requirements.
Subpart M—Enforcement
53.9000 Violations.
53.9010 Criminal penalties.
Authority: Atomic Energy Act of 1954,
secs. 11, 101, 103, 108, 122, 147, 161, 181,
182, 183, 184, 185, 186, 187, 189, 223, 234
(42 U.S.C. 2014, 2131, 2132, 2133, 2134,
2135, 2138, 2152, 2167, 2169, 2201, 2231,
2232, 2233, 2234, 2235, 2236, 2237, 2239,
2273, 2282); Energy Reorganization Act of
1974, secs. 201, 202, 206, 211 (42 U.S.C.
5841, 5842, 5846, 5851); Nuclear Waste
Policy Act of 1982, sec. 306 (42 U.S.C.
10226); National Environmental Policy Act of
1969 (42 U.S.C. 4332); 44 U.S.C. 3504 note;
Sec. 109, Pub. L. 96–295, 94 Stat. 783; Pub.
L. 115–439, 132 Stat. 5571.
§ 53.000

Purpose.

This part provides an optional
technology-inclusive, performancebased framework for the issuance,
amendment, renewal, and termination
of licenses, permits, certifications, and
approvals for commercial nuclear plants
licensed under section 103 of the
Atomic Energy Act of 1954, as amended
(the Act)(68 Stat. 919), and Title II of the
Energy Reorganization Act of 1974, as
amended (88 Stat. 1242). Also, this part
gives notice to all persons who
knowingly provide to any holder of or
applicant for an approval, certification,
permit, or license, or to a contractor,
subcontractor, or consultant of any of

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them, components, equipment,
materials, or other goods or services that
relate to the activities of a holder of or
applicant for an approval, certification,
permit, or license, subject to this part,
that they may be individually subject to
U.S. Nuclear Regulatory Commission
enforcement action for violation of the
provisions in § 53.050.
Subpart A—General Provisions
§ 53.015

Scope.

Subpart A provides general provisions
applicable to all applicants and
licensees subject to the rules of this part.

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§ 53.020

Definitions.

For the purpose of this part:
Anticipated event sequence means
event sequences expected to occur one
or more times during the life of a
commercial nuclear plant. Anticipated
event sequences take into account the
expected response of all structures,
systems, and components (SSCs) within
the plant, regardless of safety
classification.
Applicant means a person applying
for a license, permit, or other form of
Commission permission or approval
under this part.
Certified fuel handler means, for a
commercial nuclear plant, either—
(1) A non-licensed operator who has
qualified in accordance with a fuel
handler training program approved by
the Commission; or
(2) A non-licensed operator who
demonstrates compliance with the
following criteria:
(i) Has qualified in accordance with a
fuel handler training program that
demonstrates compliance with the same
requirements as training programs for
non-licensed operators required by
§ 53.830, and
(ii) Is responsible for decisions on—
(A) Safe conduct of decommissioning
activities,
(B) Safe handling and storage of spent
fuel, and
(C) Appropriate response to plant
emergencies.
Combined license (COL) means a
combined construction permit (CP) and
operating license (OL) with conditions
for a commercial nuclear plant issued
under this part.
Commercial nuclear plant means a
facility consisting of one or more
commercial nuclear reactors and
associated co-located support facilities,
including the collection of buildings,
radionuclide sources, and SSCs for
which a license, certification, or
approval is being sought under this part,
that is or will be used for producing
power for commercial electric power or

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other commercial purposes. For the
purposes of requirements in this part
that reference requirements in part 50 of
this chapter, a commercial nuclear plant
is equivalent to a nuclear power plant.
Commercial nuclear reactor means an
apparatus, other than an atomic
weapon, designed or used to sustain
nuclear fission. For the purposes of
requirements in this part that reference
requirements in 10 CFR part 50, a
commercial nuclear reactor is
equivalent to a nuclear reactor as
defined in 10 CFR 50.2.
Commission means the U.S. Nuclear
Regulatory Commission (NRC) or its
duly authorized representatives.
Consensus code or standard means
any technical standard that is—
(1) Developed or adopted by a
voluntary consensus standard body
under procedures that assure that
persons having interests within the
scope of the standard that are affected
by the provisions of the standard have
reached substantial agreement on its
adoption;
(2) Formulated in a manner that
afforded an opportunity for diverse
views to be considered; and
(3) Designated by the standards body
as a consensus code or standard.
Construction means the activities in
paragraph (1) below and does not mean
the activities in paragraph (2) below.
(1) Activities constituting
construction are those activities credited
or relied upon for demonstrating
compliance with the safety criteria
defined in subpart B of this part which
are conducted on-site to build the
commercial nuclear plant, including the
driving of piles; subsurface preparation;
placement of backfill, concrete, or
permanent retaining walls within an
excavation; installation of foundations;
or in-place assembly, erection,
fabrication, or testing, which are for—
(i) Safety-related (SR) and non-safetyrelated but safety-significant (NSRSS)
SSCs of a facility;
(ii) SSCs necessary to comply with 10
CFR part 73; or
(iii) Onsite emergency facilities
necessary to comply with § 53.855.
(2) Construction does not include—
(i) Changes for temporary use of the
land for public recreational purposes;
(ii) Site exploration, including
necessary borings to determine
foundation conditions or other
preconstruction monitoring to establish
background information related to the
suitability of the site, the environmental
impacts of construction or operation, or
the protection of environmental values;
(iii) Preparation of a site for
construction of a facility, including
clearing of the site, grading, installation

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of drainage, erosion, and other
environmental mitigation measures, and
construction of temporary roads and
borrow areas;
(iv) Erection of fences and other
access control measures;
(v) Excavation;
(vi) Erection of support buildings
(such as construction equipment storage
sheds, warehouse and shop facilities,
utilities, concrete mixing plants,
docking and unloading facilities, and
office buildings) for use in connection
with the construction of the facility;
(vii) Building of service facilities
(such as paved roads, parking lots,
railroad spurs, exterior utility and
lighting systems, potable water systems,
sanitary sewage treatment facilities, and
transmission lines);
(viii) Procurement or fabrication of
components or portions of the proposed
facility occurring at locations other than
the final, in-place location at the
facility; or
(ix) Manufacture of a nuclear power
reactor under a manufacturing license
(ML) under subpart H of this part to be
installed at the proposed site and to be
part of the proposed facility.
Custom combined license (custom
COL) means a COL that does not
reference a standard design approval or
design certification.
Decommission or decommissioning
means to remove a plant or site safely
from service and reduce residual
radioactivity to a level that permits—
(1) Release of the property for
unrestricted use and termination of the
license; or
(2) Release of the property under
restricted conditions and termination of
the license.
Defense in depth means inclusion of
two or more independent and
redundant layers of defense in the
design of a facility and its operating
procedures to compensate for
uncertainties such that no single layer of
defense, no matter how robust, is
exclusively relied upon. Defense in
depth includes, but is not limited to, the
use of access controls, physical barriers,
redundant and diverse safety functions,
and emergency response measures.
Design-basis accidents (DBAs) means
postulated event sequences that are
used to set functional design criteria
and performance objectives for the
design of SR SSCs through deterministic
analyses. Design-basis accidents are a
type of licensing-basis event and are
based on the capabilities and
reliabilities of SR SSCs needed to
mitigate and prevent event sequences,
respectively.
Design-basis external hazard level
means the level of severity or intensity

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of an external hazard for which the SR
SSCs are protected against or designed
to withstand without losing their
capability to perform their safety
functions.
Design features means the active and
passive SSCs and the inherent
characteristics of those SSCs that
contribute to limiting the total effective
dose equivalent to individual members
of the public during normal operations
and prevent or mitigate the
consequences of event sequences.
Electric utility means any entity that
generates or distributes electricity and
that recovers the cost of this electricity,
either directly or indirectly, through
rates established by the entity itself or
by a separate regulatory authority.
Investor-owned utilities, including
generation or distribution subsidiaries,
public utility districts, municipalities,
rural electric cooperatives, and State
and Federal agencies, including
associations of any of the foregoing, are
included within the meaning of
‘‘electric utility.’’
Event sequence means a postulated
initiating event defined for a set of
initial plant conditions followed by
system, safety function, and operator
successes or failures, and terminating in
a specified end state depending on the
system, safety function, and operator
successes and failures (e.g., prevention
of release of radioactive material or
release in one of the reactor-specific
release categories). An event sequence
may include many unique variations of
events that are similar in terms of
results or end states.
Exclusion area means that area
surrounding the reactor, in which the
reactor licensee has the authority to
determine all activities including
exclusion or removal of personnel and
property from the area. This area may be
traversed by a highway, railroad, or
waterway, provided these are not so
close to the facility as to interfere with
normal operations of the facility and
provided appropriate and effective
arrangements are made to control traffic
on the highway, railroad, or waterway,
in case of emergency, to protect the
public health and safety. Residence
within the exclusion area must normally
be prohibited. In any event, residents
must be subject to ready removal in case
of necessity. Activities unrelated to
operation of the reactor may be
permitted in an exclusion area under
appropriate limitations, provided that
no significant hazards to the public
health and safety will result.
Fission product release means the
amount and composition of radioactive
material released to the environment,
after accounting for any retention of

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radionuclides provided by reactor
design features.
Fuel means special nuclear material
(SNM) or source material, discrete
elements that physically contain SNM
or source material, and homogeneous
mixtures that contain SNM or source
material, intended to or used to create
power in a commercial nuclear plant.
Functional design criteria means
metrics for the performance of SSCs. For
SR SSCs, these criteria define
performance metrics necessary to
demonstrate compliance with the safety
criteria in § 53.210. For NSRSS SSCs,
these criteria define performance
metrics necessary to demonstrate
compliance with the safety criteria in
§ 53.220.
License, when used in the context of
a facility, means a limited work
authorization, CP, OL, early site permit,
COL, or ML under this part, or a
renewed license issued by the
Commission under this part. When used
in the context of a license authorizing
an individual to manipulate the controls
of a facility, license means a license
issued by the Commission to perform
the function of an operator, senior
operator, or generally licensed reactor
operator as defined in this part.
Licensee means a person who is
authorized to conduct activities under a
license issued under this part by the
Commission.
Licensing-basis events means a
collection of event sequences
considered in the design and licensing
of the commercial nuclear plant.
Licensing-basis events are unplanned
events and include anticipated event
sequences, unlikely event sequences,
very unlikely event sequences, and
DBAs.
Licensing-basis information means the
information contained in regulations,
orders, licenses, certifications, or
approvals issued by the NRC for a
commercial nuclear plant licensed
under this part and that information
submitted to the NRC by an applicant or
licensee in a Safety Analysis Report,
program description, or other licensingrelated document required under this
part.
Low-population zone means the area
immediately surrounding the exclusion
area which contains residents, the total
number and density of which are such
that there is a reasonable probability
that appropriate protective measures
could be taken on their behalf in the
event of a serious accident. A
permissible population density or total
population within this zone is not
included in this definition because the
situation may vary from case to case.
Whether a specific number of people

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can, for example, be evacuated from a
specific area or instructed to take shelter
on a timely basis, will depend on many
factors such as location, number and
size of highways, scope and extent of
advance planning, and actual
distribution of residents within the area.
Major decommissioning activity
means, for a commercial nuclear plant,
any activity that results in permanent
removal of major radioactive
components, permanently modifies the
structure of the containment, if
applicable, or results in dismantling
components for shipment containing
greater than class C waste in accordance
with 10 CFR 61.55.
Major feature of the emergency plans
means an aspect of those plans
necessary to:
(1) Address in whole or part either
one or more of the 16 standards in 10
CFR 50.47(b) or the requirements of 10
CFR 50.160(b), as applicable; or
(2) Describe the emergency planning
zones as required in § 53.1109(g).
Manufactured reactor means the
essential portions of a nuclear reactor
that are manufactured under an ML and
subsequently transported and
incorporated into a commercial nuclear
plant under a COL.
Manufacturing license means a
license issued under this part that
authorizes the manufacture of
manufactured reactors but not its
construction, installation, or operation.
Non-Safety-Related but SafetySignificant (NSRSS) SSCs means those
SSCs which are not SR but are relied on
to achieve adequate defense in depth or
perform risk-significant functions and
warrant special treatment.
Non-Safety-Significant SSCs means
those SSCs that are not SR or NSRSS,
are not relied on to achieve adequate
defense in depth or to perform risksignificant functions, and do not
warrant special treatment.
Person means—
(1) any individual, corporation,
partnership, firm, association, trust,
estate, public or private institution,
group, government agency other than
the Commission or the Department,
except that the Department shall be
considered a person to the extent that its
facilities are subject to the licensing and
related regulatory authority of the
Commission pursuant to section 202 of
the ERA, any State or any political
subdivision of, or any political entity
within a State, any foreign government
or nation or any political subdivision of
any such government or nation, or other
entity; and
(2) any legal successor, representative,
agent, or agency of the foregoing.

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Population center distance means the
distance from the reactor to the nearest
boundary of a densely populated center
containing more than about 25,000
residents.
Probabilistic risk assessment means a
quantitative assessment of the risk
associated with plant operation and
maintenance that is measured in terms
of event sequence occurrence
frequencies and consequences.
Programmatic controls means
administrative procedures that govern
human action in implementing
programs and operating, monitoring,
and maintaining SSCs and equipment of
a commercial nuclear plant.
Programmatic controls considered to be
licensing basis information are specified
in an application for a requested activity
of the Commission.
Quality assurance (QA) means all
those planned and systematic actions
necessary to ensure that a structure,
system, or component will perform
satisfactorily in service. Quality
assurance includes quality control,
which comprises those QA actions
related to the physical characteristics of
a material, structure, component, or
system which provide a means to
control the quality of the material,
structure, component, or system to
predetermined requirements.
Safety criteria means performancebased metrics that establish a level of
safety provided in requirements in
§§ 53.210 and 53.220.
Safety-related structures, systems, or
components means those SSCs that are
relied upon to demonstrate compliance
with the safety criteria in § 53.210 and
warrant special treatment.
Small modular reactor means a power
reactor, which may be of modular
design as defined in 10 CFR 52.1,
licensed under this part to produce heat
energy up to 1,000 megawatts thermal
per module.
Site characteristics means the actual
physical, environmental, and
demographic features of a site. Site
characteristics are specified in an early
site permit or in a Preliminary or Final
Safety Analysis Report for a limited
work authorization, CP, or COL, as
applicable.
Site parameters are the postulated
physical, environmental, and
demographic features of an assumed
site. Site parameters are specified in a
standard design approval, standard
design certification, or ML.
Source material means source
material as defined in subsection 11z. of
the Atomic Energy Act of 1954, as
amended, (the Act) and in the
regulations contained in part 40 of this
chapter.

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Special nuclear material (SNM)
means:
(1) Plutonium, uranium-233, uranium
enriched in the isotope-233 or in the
isotope-235, and any other material
which the Commission, pursuant to the
provisions of section 51 of the Act,
determines to be SNM, but does not
include source material; or
(2) Any material artificially enriched
by any of the foregoing, but does not
include source material.
Special treatment means those
requirements, such as QA and
programmatic controls, that ensure that
SR and NSRSS SSCs will provide
defense in depth or perform risksignificant functions. The requirements
also ensure that the SSCs will perform
under the service conditions and with
the reliability assumed in the analysis
performed under § 53.450 to
demonstrate compliance with the safety
criteria in §§ 53.210 and 53.220.
Standard design means a design
which is sufficiently detailed and
complete to support certification or
approval in accordance with subpart H
of this part, and which is usable under
of this part for a multiple number of
units or at a multiple number of sites
without reopening or repeating the
review.
Standard design approval or design
approval means an NRC staff approval,
issued under subpart H of this part, of
a final standard design for a commercial
nuclear plant. The approval may be for
either the final design for the entire
reactor facility or the final design of
major portions thereof.
Standard design certification or
design certification means a
Commission approval, issued under
subpart H of this part, of a final standard
design for a nuclear power facility. This
design may be referred to as a certified
standard design.
Total effective dose equivalent means
the sum of the effective dose equivalent
(for external exposures) and the
committed effective dose equivalent (for
internal exposures).
Utilization facility means any
commercial nuclear reactor other than
one designed or used primarily for the
formation of plutonium or uranium-233.
Unlikely event sequences means event
sequences that are not expected to occur
in the life of a commercial nuclear plant
and are less likely than anticipated
event sequences, but are infrequent
rather than rare. Unlikely event
sequences take into account the
expected response of all SSCs within
the plant regardless of safety
classification.
Very unlikely event sequences means
event sequences that are not expected to

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occur in the life of a commercial nuclear
plant, are less likely than an unlikely
event sequence, and are rare. Very
unlikely event sequences take into
account the expected response of all
SSCs within the plant regardless of
safety classification.
§ 53.030

[Reserved]

§ 53.040

Written communications.

(a) General requirements. All
correspondence, reports, applications,
and other written communications from
the applicant or licensee to the NRC
concerning the regulations in this part
or individual license conditions must be
sent either by mail addressed: ATTN:
Document Control Desk, U.S. Nuclear
Regulatory Commission, Washington,
DC 20555–0001; by hand delivery to the
NRC’s offices at 11555 Rockville Pike,
Rockville, Maryland, between the hours
of 8:15 a.m. and 4 p.m. eastern time; or,
where practicable, by electronic
submission, for example, via Electronic
Information Exchange, email, or CD–
ROM. Electronic submissions must be
made in a manner that enables the NRC
to receive, read, authenticate, distribute,
and archive the submission, and process
and retrieve it a single page at a time.
Detailed guidance on making electronic
submissions can be obtained by visiting
the NRC’s website at http://
www.nrc.gov/site-help/esubmittals.html; by email to
MSHD.Resource@nrc.gov; or by writing
the Office of the Chief Information
Officer, U.S. Nuclear Regulatory
Commission, Washington, DC 20555–
0001. The guidance discusses, among
other topics, the formats the NRC can
accept, the use of electronic signatures,
and the treatment of nonpublic
information. If the communication is on
paper, the signed original must be sent.
If a submission due date falls on a
Saturday, Sunday, or Federal holiday,
the next Federal working day becomes
the official due date.
(b) Distribution requirements. Copies
of all correspondence, reports, and other
written communications concerning the
regulations in this part or individual
license conditions, or the terms and
conditions of an early site permit or
standard design approval, must be
submitted to the persons listed below
(addresses for the NRC Regional Offices
are listed in appendix D to 10 CFR part
20).
(1) Applications for amendment of
permits and licenses, reports, and other
communications. All written
communications (including responses to
generic letters, bulletins, information
notices, regulatory information
summaries, inspection reports, and

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miscellaneous requests for additional
information) that are required of holders
of licenses, permits, and design
approvals issued pursuant to this part,
must be submitted as follows, except as
otherwise specified in paragraphs (b)(2)
through (7) of this section: to the NRC’s
Document Control Desk (if on paper, the
signed original), with a copy to the
appropriate Regional Office, and a copy
to the appropriate NRC Resident
Inspector if one has been assigned to the
site of the facility or the place of
manufacture of a reactor licensed under
this part.
(2) Applications for permits and
licenses, and amendments to
applications. Applications for licenses,
permits, and design approvals and
amendments to any of these types of
applications must be submitted to the
NRC’s Document Control Desk, with a
copy to the appropriate Regional Office,
and a copy to the appropriate NRC
Resident Inspector if one has been
assigned to the facility or the place of
manufacture of a reactor licensed under
this part, except as otherwise specified
in paragraphs (b)(3) through (9) of this
section. If the application or amendment
is on paper, the submission to the
Document Control Desk must be the
signed original.
(3) Acceptance review application.
Written communications required for an
application for determination of
suitability for docketing must be
submitted to the NRC’s Document
Control Desk, with a copy to the
appropriate Regional Office. If the
communication is on paper, the
submission to the Document Control
Desk must be the signed original.
(4) Security plan and related
submissions. Written communications,
as defined in paragraphs (b)(4)(i)
through (v) of this section, must be
submitted to the NRC’s Document
Control Desk, with a copy to the
appropriate Regional Office. If the
communication is on paper, the
submission to the Document Control
Desk must be the signed original.
Submissions should include the
following as appropriate:
(i) Physical security plan;
(ii) Safeguards contingency plan;
(iii) Cybersecurity plan;
(iv) Change to security plan, guard
training and qualification plan,
safeguards contingency plan, or
cybersecurity plan made without prior
Commission approval under § 53.1565;
and
(v) Application for amendment of
physical security plan, guard training
and qualification plan, safeguards
contingency plan, or cybersecurity plan
under § 53.1510.

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(5) Emergency plan and related
submissions. Written communications
as defined in paragraphs (b)(5)(i)
through (iii) of this section must be
submitted to the NRC’s Document
Control Desk, with a copy to the
appropriate Regional Office, and a copy
to the appropriate NRC Resident
Inspector if one has been assigned to the
site of the facility. If the communication
is on paper, the submission to the
Document Control Desk must be the
signed original. Submissions should
include the following as appropriate:
(i) Emergency plan;
(ii) Change to an emergency plan
under § 53.1565; and
(iii) Emergency implementing
procedures under § 53.855.
(6) Updated Final Safety Analysis
Report. An Updated Final Safety
Analysis Report or replacement pages
under § 53.1545 must be submitted to
the NRC’s Document Control Desk, with
a copy to the appropriate Regional
Office, and a copy to the appropriate
NRC Resident Inspector if one has been
assigned to the site of the facility or the
place of manufacture of a reactor
licensed under this part. Paper copy
submissions may be made using
replacement pages; however, if a
licensee chooses to use electronic
submission, all subsequent updates or
submissions must be performed
electronically on a total replacement
basis. If the communication is on paper,
the submission to the Document Control
Desk must be the signed original. If the
communications are submitted
electronically, see Guidance for
Electronic Submissions to the
Commission.
(7) Quality assurance related
submissions. (i) A change to the Safety
Analysis Report QA program
description under § 53.1565, or a change
to a licensee’s NRC-accepted QA topical
report under § 53.1565, must be
submitted to the NRC’s Document
Control Desk, with a copy to the
appropriate Regional Office, and a copy
to the appropriate NRC Resident
Inspector if one has been assigned to the
site of the facility or the place of
manufacture of a reactor licensed under
this part. If the communication is on
paper, the submission to the Document
Control Desk must be the signed
original.
(ii) A change to an NRC-accepted QA
topical report from non-licensees (i.e.,
architect/engineers, nuclear steam
supply system suppliers, fuel suppliers,
constructors, etc.) must be submitted to
the NRC’s Document Control Desk. If
the communication is on paper, the
signed original must be sent.

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(8) Certification of permanent
cessation of operations. The licensee’s
certification of permanent cessation of
operations, under subpart G of this part,
must state the date on which operations
have ceased or will cease, and must be
submitted to the NRC’s Document
Control Desk. This submission must be
under oath or affirmation.
(9) Certification of permanent fuel
removal. The licensee’s certification of
permanent fuel removal, under subpart
G of this part, must state the date on
which the fuel was removed from the
reactor vessel and the disposition of the
fuel, and must be submitted to the
NRC’s Document Control Desk. This
submission must be under oath or
affirmation.
(c) Form of communications. All
paper copies submitted to demonstrate
compliance with the requirements set
forth in paragraph (b) of this section
must be typewritten, printed, or
otherwise reproduced in permanent
form on unglazed paper. Exceptions to
these requirements imposed on paper
submissions may be granted for the
submission of micrographic,
photographic, or similar forms.
(d) Regulation governing submission.
Licensees, applicants, and holders of
standard design approvals submitting
correspondence, reports, and other
written communications under the
regulations of this part are requested but
not required to cite whenever practical,
in the upper right corner of the first
page of the submission, the specific
regulation or other basis requiring
submission.
§ 53.050

Deliberate misconduct.

(a) Any licensee or applicant for a
license; holder of or applicant for a
standard design approval; applicant for
a standard design certification;
employee of a licensee, holder of a
standard design approval, or applicant
for a license, standard design approval,
or standard design certification; or any
contractor (including a supplier or
consultant), subcontractor, employee of
a contractor or subcontractor of any
licensee or applicant for a license,
holder of or applicant for a standard
design approval, or applicant for a
standard design certification, who
knowingly provides to any licensee,
applicant, contractor, or subcontractor,
any components, equipment, materials,
or other goods or services that relate to
a licensee’s or applicant’s activities in
this part, may not—
(1) Engage in deliberate misconduct
that causes or would have caused, if not
detected, a licensee or applicant to be in
violation of any rule, regulation, or
order; or any term, condition, or

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limitation of any license issued by the
Commission; or
(2) Deliberately submit to the NRC, a
licensee, an applicant, or a licensee’s or
applicant’s contractor or subcontractor,
information that the person submitting
the information knows to be incomplete
or inaccurate in some respect material to
the NRC.
(b) A person who violates paragraph
(a)(1) or (2) of this section may be
subject to enforcement action in
accordance with the procedures in
subpart B of 10 CFR part 2.
(c) For the purposes of paragraph
(a)(1) of this section, deliberate
misconduct by a person means an
intentional act or omission that the
person knows—
(1) Would cause a licensee or
applicant to be in violation of any rule,
regulation, or order; or any term,
condition, or limitation, of any license
issued by the Commission; or
(2) Constitutes a violation of a
requirement, procedure, instruction,
contract, purchase order, or policy of a
licensee, applicant, contractor, or
subcontractor.

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§ 53.060

Employee protection.

(a) Discrimination by a Commission
licensee, holder of a standard design
approval, an applicant for a license,
standard design certification, or
standard design approval, a contractor
or subcontractor of a Commission
licensee, holder of a standard design
approval, applicant for a license,
standard design certification, or
standard design approval, against an
employee for engaging in certain
protected activities is prohibited.
Discrimination includes discharge and
other actions that relate to
compensation, terms, conditions, or
privileges of employment. The protected
activities are established in section 211
of the Energy Reorganization Act of
1974, as amended, and in general are
related to the administration or
enforcement of a requirement imposed
under the Act or the Energy
Reorganization Act of 1974, as
amended.
(1) The protected activities include
but are not limited to—
(i) Providing the Commission or his or
her employer information about alleged
violations of either of the statutes
named in paragraph (a) of this section
or possible violations of requirements
imposed under either of those statutes;
(ii) Refusing to engage in any practice
made unlawful under either of the
statutes named in paragraph (a) of this
section or under these requirements if
the employee has identified the alleged
illegality to the employer;

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(iii) Requesting the NRC to institute
action against his or her employer for
the administration or enforcement of
these requirements;
(iv) Testifying in any Commission
proceeding, or before Congress, or at any
Federal or State proceeding regarding
any provision (or proposed provision) of
either of the statutes named in
paragraph (a) of this section; and
(v) Assisting or participating in, or
being about to assist or participate in,
these activities.
(2) These activities are protected even
if no formal proceeding is actually
initiated as a result of the employee
assistance or participation.
(3) This section has no application to
any employee alleging discrimination
prohibited by this section who, acting
without direction from his or her
employer (or the employer’s agent),
deliberately causes a violation of any
requirement of the Energy
Reorganization Act of 1974, as
amended, or the Act.
(b) Any employee who believes that
they have been discharged or otherwise
discriminated against by any person for
engaging in protected activities
specified in paragraph (a)(1) of this
section may seek a remedy for the
discharge or discrimination through an
administrative proceeding in the
Department of Labor. The
administrative proceeding must be
initiated within 180 days after an
alleged violation occurs. The employee
may do this by filing a complaint
alleging the violation with the
Department of Labor, Wage and Hour
Division. The Department of Labor may
order reinstatement, back pay, and
compensatory damages.
(c) A violation of paragraph (a), (e), or
(f) of this section by a Commission
licensee, a holder of a standard design
approval, an applicant for a Commission
license, standard design certification, or
a standard design approval, or a
contractor or subcontractor of a
Commission licensee, holder of a
standard design approval, or any
applicant may be grounds for—
(1) Denial, revocation, or suspension
of the license or standard design
approval;
(2) Withdrawal or revocation of a
proposed or final standard design
certification;
(3) Imposition of a civil penalty on the
licensee, holder of a standard design
approval, or applicant (including an
applicant for a standard design
certification under this part following
Commission adoption of final design
certification rule) or a contractor or
subcontractor of the licensee, holder of

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a standard design approval, or
applicant; or
(4) Other enforcement action.
(d) Actions taken by an employer, or
others, which adversely affect an
employee may be predicated upon
nondiscriminatory grounds. The
prohibition applies when the adverse
action occurs because the employee has
engaged in protected activities. An
employee’s engagement in protected
activities does not automatically render
him or her immune from discharge or
discipline for legitimate reasons or from
adverse action dictated by
nonprohibited considerations.
(e)(1) Each holder or applicant for a
license or design approval, must
prominently post the revision of NRC
Form 3, ‘‘Notice to Employees,’’
referenced in § 19.11(e)(1) of this
chapter. This form must be posted at
locations sufficient to permit employees
protected by this section to observe a
copy on the way to or from their place
of work. Premises must be posted no
later than 30 days after an application
is docketed and remain posted while the
application is pending before the
Commission, during the term of the
license, and for 30 days following
license termination.
(2) Copies of NRC Form 3 may be
obtained by writing to the Regional
Administrator of the appropriate NRC
Regional Office listed in appendix D to
10 CFR part 20, via email to
Forms.Resource@nrc.gov, or by visiting
the NRC’s online library at http://
www.nrc.gov/reading-rm/doccollections/forms/.
(f) No agreement affecting the
compensation, terms, conditions, or
privileges of employment, including an
agreement to settle a complaint filed by
an employee with the Department of
Labor pursuant to section 211 of the
Energy Reorganization Act of 1974, as
amended, may contain any provision
which would prohibit, restrict, or
otherwise discourage an employee from
participating in protected activity as
defined in paragraph (a)(1) of this
section including, but not limited to,
providing information to the NRC or to
his or her employer on potential
violations or other matters within NRC’s
regulatory responsibilities.
(g) Part 19 of 10 CFR sets forth
requirements and regulatory provisions
applicable to licensees, holders of a
standard design approval, applicants for
a license, standard design certification,
or standard design approval, and
contractors or subcontractors of a
Commission licensee, or holder of a
standard design approval, and are in
addition to the requirements in this
section.

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§ 53.070 Completeness and accuracy of
information.

(a) Information provided to the
Commission by a holder of a license,
permit, design certification, or standard
design approval under this part or an
applicant for a license, permit, design
certification, or standard design
approval under this part, and
information required by statute or by the
Commission’s regulations, orders,
license conditions, or terms and
conditions of a standard design
approval to be maintained by the
applicant or the licensee must be
complete and accurate in all material
respects.
(b) Each applicant or licensee, each
holder of a standard design approval
under this part, and each applicant for
a standard design certification under
this part following Commission
adoption of a final design certification
regulation, must notify the Commission
of information identified by the
applicant or licensee as having for the
regulated activity a significant
implication for public health and safety
or common defense and security. An
applicant, licensee, or holder violates
this paragraph only if the applicant,
licensee, or holder fails to notify the
Commission of information that the
applicant, licensee, or holder has
identified as having a significant
implication for public health and safety
or common defense and security.
Notification must be provided to the
Administrator of the appropriate
Regional Office within 2 working days
of identifying the information. This
requirement is not applicable to
information which is already required to
be provided to the Commission by other
reporting or updating requirements.

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§ 53.080

Specific exemptions.

(a) The Commission may, upon
application by any interested person or
upon its own initiative, grant
exemptions from the requirements of
the regulations of this part, which are
authorized by law, will not present an
undue risk to the public health and
safety, and are consistent with the
common defense and security.
(b) The Commission will not consider
granting an exemption unless special
circumstances are present. Special
circumstances are present whenever—
(1) Application of the regulation in
the particular circumstances conflicts
with other rules or requirements of the
Commission;
(2) Application of the regulation in
the particular circumstances would not
serve the underlying purpose of the rule
or is not necessary to achieve the
underlying purpose of the rule;

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(3) Compliance would result in undue
hardship or other costs that are
significantly in excess of those
contemplated when the regulation was
adopted, or that are significantly in
excess of those incurred by others
similarly situated;
(4) The exemption would result in
benefit to the public health and safety
that compensates for any decrease in
safety that may result from the grant of
the exemption;
(5) The exemption would provide
only temporary relief from the
applicable regulation and the licensee or
applicant has made good faith efforts to
comply with the regulation; or
(6) There is present any other material
circumstance not considered when the
regulation was adopted for which it
would be in the public interest to grant
an exemption. If such condition is relied
on exclusively for demonstrating
compliance with paragraph (b) of this
section, the exemption may not be
granted until the Executive Director for
Operations has consulted with the
Commission.
(c) Any person may request an
exemption permitting the conduct of
construction activities prior to the
issuance of a CP. The Commission may
grant such an exemption upon
considering and balancing the following
factors:
(1) Whether conduct of the proposed
activities will give rise to a significant
adverse impact on the environment and
the nature and extent of such impact, if
any;
(2) Whether redress of any adverse
environment impact from conduct of the
proposed activities can reasonably be
effective should such redress be
necessary;
(3) Whether conduct of the proposed
activities would foreclose subsequent
adoption of alternatives; and
(4) The effect of delay in conducting
such activities on the public interest,
including whether the power needs to
be used by the proposed facility, the
availability of alternative sources, if any,
to meet those needs on a timely basis
and delay costs to the applicant and to
consumers.
(d) Issuance of such an exemption
must not be deemed to constitute a
commitment to issue a CP. During the
period of any exemption granted
pursuant to paragraph (c) of this section,
any activities conducted must be carried
out in such a manner as will minimize
or reduce their environmental impact.
(e) The Commission’s consideration of
requests for exemptions from
requirements of the regulations of other
parts in this chapter that are applicable
by virtue of this part must be governed

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by the exemption requirements of those
parts.
§ 53.090

Standards for review.

(a) Common standards. In
determining that a CP, OL, early site
permit, COL, or ML under this part will
be issued to an applicant, the
Commission will be guided by the
following considerations:
(1) Except for an early site permit or
ML, the processes to be performed, the
operating procedures, the facility and
equipment, the use of the facility, and
other technical specifications, or the
proposals, in regard to any of the
foregoing, collectively provide
reasonable assurance that the applicant
will comply with the regulations in this
chapter, including the regulations in 10
CFR part 20, and that the health and
safety of the public will not be
endangered.
(2) The applicant for a CP, OL, COL,
or ML is technically and financially
qualified to engage in the proposed
activities in accordance with the
regulations in this chapter. However, no
consideration of financial qualification
is necessary for an electric utility
applicant for an OL for a utilization
facility of the type described in
paragraph (d) of this section or for an
applicant for an ML.
(3) The issuance of a CP, OL, early site
permit, COL, or ML to the applicant will
not, in the opinion of the Commission,
be inimical to the common defense and
security or to the health and safety of
the public.
(4) Any applicable requirements of
subpart A of 10 CFR part 51 have been
satisfied.
(b) Additional standards for licenses.
In determining whether a license will be
issued to an applicant, the Commission
will, in addition to applying the
standards set forth in paragraph (a) of
this section, consider whether the
proposed activities will serve a useful
purpose proportionate to the quantities
of SNM or source material to be utilized.
(c) Additional standards and
provisions affecting licenses for
commercial power. In addition to
applying the standards set forth in
paragraphs (a) and (b) of this section,
paragraphs (c)(1) through (c)(4) of this
section apply in the case of a license for
a facility for the generation of
commercial power.
(1) The NRC will—
(i) Give notice in writing of each
application to the regulatory agency or
State as may have jurisdiction over the
rates and services incident to the
proposed activity;
(ii) Publish notice of the application
in trade or news publications as it

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deems appropriate to give reasonable
notice to municipalities, private
utilities, public bodies, and cooperatives
which might have a potential interest in
the utilization or production facility;
and
(iii) Publish notice of the application
once each week for four consecutive
weeks in the Federal Register. No
license will be issued by the NRC prior
to the giving of these notices and until
four weeks after the last notice is
published in the Federal Register.
(2) If there are conflicting applications
for a limited opportunity for such
license, the Commission will give
preferred consideration in the following
order: first, to applications submitted by
public or cooperative bodies for
facilities to be located in high cost
power areas in the United States;
second, to applications submitted by
others for facilities to be located in such
areas; third, to applications submitted
by public or cooperative bodies for
facilities to be located in areas other
than high cost power areas; and, fourth,
to all other applicants.
(3) The licensee who transmits
electric energy in interstate commerce,
or sells it at wholesale in interstate
commerce, must be subject to the
regulatory provisions of the Federal
Power Act.
(4) Nothing shall preclude any
government agency, now or hereafter
authorized by law to engage in the
production, marketing, or distribution of
electric energy, if otherwise qualified,
from obtaining a CP, OL, or COL under
this part for a utilization facility for the
primary purpose of producing electric
energy for disposition for ultimate
public consumption.
(d) Licenses for commercial nuclear
plants. A license will be issued, to an
applicant who qualifies, for any one or
more of the following: to transfer or
receive in interstate commerce, or
manufacture, produce, transfer, acquire,
possess, or use a utilization facility for
industrial or commercial purposes.
§ 53.100

Jurisdictional limits.

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No permit, license, standard design
approval, or standard design
certification under this part shall be
deemed to have been issued for
activities that are not under or within
the jurisdiction of the United States.
§ 53.110

Attacks and destructive acts.

Licensees, applicants for licenses,
permits, certifications, and design
approvals, and applicants for an
amendment to any license, permit,
certification, or design approval under
this part are not required to provide for
design features or other measures for the

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specific purpose of protection against
the effects of—
(a) Attacks and destructive acts,
including sabotage, directed against the
facility by an enemy of the United
States, whether a foreign government or
other person; or
(b) Use or deployment of weapons
incident to U.S. defense activities.
§ 53.115 Rights related to special nuclear
material.

(a) No right to the SNM will be
conferred by a license issued under this
part except as may be defined by the
license.
(b) Neither a license issued under this
part, nor any right thereunder, nor any
right to utilize or produce SNM may be
transferred, assigned, or disposed of in
any manner, either voluntarily or
involuntarily, directly or indirectly,
through transfer of control of the license
to any person, unless the Commission,
after securing full information, finds
that the transfer is in accordance with
the provisions of the Act and gives its
consent in writing.
§ 53.117 License suspension and rights of
recapture.

Any license issued under this part
must be subject to suspension and to the
rights of recapture of the material or
control of the facility reserved to the
Commission under section 108 of the
Act in a state of war or national
emergency declared by Congress.
§ 53.120 Information collection
requirements: OMB approval.

(a) The NRC has submitted the
information collection requirements
contained in this part to the Office of
Management and Budget (OMB) for
approval as required by the Paperwork
Reduction Act (44 U.S.C. 3501 et seq.).
The NRC may not conduct or sponsor,
and a person is not required to respond
to, a collection of information unless it
displays a currently valid OMB control
number. OMB has approved the
information collection requirements
contained in this part under control
number 3150–XXXX.
(b) The approved information
collection requirements contained in
this part appear in §§ 53.070, 53.080,
53.240, 53.410, 53.420, 53.425, 53.430,
53.440, 53.450, 53.480, 53.500, 53.540,
53.605, 53.610, 53.620, 53.700, 53.710,
53.715, 53.720, 53.730, 53.780, 53.785,
53.805, 53.810, 53.815, 53.830, 53.850,
53.855, 53.865, 53.870, 53.875, 53.880,
53.910, 53.1010, 53.1020, 53.1030,
53.1045, 53.1060, 53.1070, 53.1075,
53.1080, 53.1100, 53.1109, 53.1115,
53.1130, 53.1140, 53.1144, 53.1146,
53.1173, 53. 1182, 53.1188, 53.1200,
53.1206, 53.1209, 53.1210, 53.1221,

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53.1230, 53.1236, 53.1239, 53.1241,
53.1254, 53.1257, 53,1263, 53.1270,
53.1276, 53.1279, 53.1282, 53.1288,
53.1295, 53.1300, 53.1306, 53.1309,
53.1312, 53.1327, 53.1330, 53.1333,
53.1336, 53.1348, 53.1360, 53.1366,
53.1369, 53.1372, 53.1384, 53.1410,
53.1413, 53.1416, 53.1419, 53.1437,
53.1449, 53.1452, 53.1458, 53.1470,
53.1505, 53.1510, 53.1515, 53.1525,
53.1530, 53.1535, 53.1540, 53.1545,
53.1550, 53.1560, 53.1565, 53.1570,
53.1575, 53.1580, 53.1620, 53.1630,
53.1645, 53.1680, 53.1690, 53.1720.
(c) This part contains information
collection requirements in addition to
those approved under the control
number specified in paragraph (a) of
this section. The information collection
requirement and the control numbers
under which it is approved are as
follows:
(1) In §§ 53.765, 53.770, 53.780, and
53.795, NRC Form 396 is approved
under control number 3150–0024.
(2) In §§ 53.775 and 53.795, NRC
Form 398 is approved under control
number 3150–0090.
(3) In § 53.1640, NRC Form 366 is
approved under control number 3150–
0104.
(4) In § 53.1630, NRC Form 361 is
approved under control number 3150–
0238.
(5) In § 53.1650, International Atomic
Energy Agency Design Information
Questionnaire forms are approved under
control number 3150–0056.
(6) In § 53.1650, DOC/NRC Form AP–
A and associated forms are approved
under control numbers 0694–0135.
Subpart B—Technology-Inclusive
Safety Requirements
§ 53.210 Safety criteria for design-basis
accidents.

Design features and programmatic
controls must be provided for each
commercial nuclear plant such that
identification and analyses of designbasis accidents (DBAs) in accordance
with § 53.240 demonstrate the
following:
(a) An individual located at any point
on the boundary of the exclusion area
for any 2-hour period following the
onset of the postulated fission product
release would not receive a radiation
dose in excess of 25 rem (250
millisieverts) total effective dose
equivalent (TEDE); and
(b) An individual located at any point
on the outer boundary of the lowpopulation zone who is exposed to the
radioactive cloud resulting from the
postulated fission product release
(during the entire period of its passage)
would not receive a radiation dose in

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excess of 25 rem (250 millisieverts)
TEDE.1
1 The use of 25 rem TEDE is not intended
to imply that this number constitutes an
acceptable limit for an emergency dose to the
public under accident conditions. Rather,
this dose value has been set forth in this
section as a reference value, which can be
used in the evaluation of plant design
features with respect to postulated reactor
accidents, to assure that these designs
provide assurance of low risk of public
exposure to radiation, in the event of an
accident.

§ 53.220 Safety criteria for licensing-basis
events other than design-basis accidents.

Design features and programmatic
controls must be provided for each
commercial nuclear plant such that
identification and analysis of licensingbasis events (LBEs) other than DBAs in
accordance with § 53.240 demonstrate
the following:
(a) Plant SSCs, personnel, and
programs provide the necessary
capabilities and maintain the necessary
reliability to address LBEs other than
DBAs in accordance with §§ 53.240 and
53.450(e), and provide measures for
defense in depth in accordance with
§ 53.250; and
(b) The analysis of risks to public
health and safety resulting from LBEs
other than DBAs under § 53.450(e)
includes comprehensive risk metrics
that satisfy associated risk performance
objectives that are acceptable to the NRC
and provide an appropriate level of
safety.

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§ 53.230

Safety functions.

(a) The primary safety function is
limiting the release of radioactive
materials from the facility and must be
maintained during normal operation
and for LBEs over the life of the plant.
(b) Additional safety functions needed
to support the retention of radioactive
materials during LBEs—such as
controlling reactivity, heat generation,
heat removal, and chemical
interactions—must be identified for
each commercial nuclear plant.
(c) The primary and additional safety
functions are required to satisfy the
safety criteria defined in §§ 53.210 and
53.220, or more restrictive alternative
criteria adopted under § 53.470, and
must be fulfilled by the design features,
human actions, and programmatic
controls specified throughout this part.
§ 53.240

Licensing-basis events.

(a) Licensing-basis events must be
identified for each commercial nuclear
plant and analyzed under § 53.450 to
demonstrate that the safety
requirements in this subpart have been
satisfied.

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(b) The identified LBEs, ranging from
anticipated event sequences to very
unlikely event sequences, must
collectively address combinations of
malfunctions of plant SSCs, human
errors, facility hazards, and the effects of
external hazards.
(c) The analysis of LBEs must—
(1) Include analysis of one or more
DBAs under § 53.450(f);
(2) Confirm the adequacy of design
features and programmatic controls
needed to satisfy the safety criteria
defined in §§ 53.210 and 53.220, or
more restrictive alternative criteria
adopted under § 53.470, and
(3) Establish related functional
requirements for plant SSCs, personnel,
and programs.
§ 53.250

Defense in depth.

(a) Measures must be taken for each
commercial nuclear plant to ensure
appropriate defense in depth is
provided to compensate for
uncertainties in the analysis of the
safety criteria such that there is
reasonable assurance that the safety
criteria in this subpart are met over the
life of the plant.
(b) The uncertainties that must be
addressed under paragraph (a) of this
section include those related to the state
of knowledge and modeling capabilities,
the ability of barriers to limit the release
of radioactive materials from the facility
during LBEs other than DBAs, the
reliability and performance of plant
SSCs and personnel, and the
effectiveness of programmatic controls.
(c) The safety analysis may not rely
upon a single engineered design feature,
human action, or programmatic control,
no matter how robust, to address the
range of LBEs other than DBAs.
§ 53.260

Normal operations.

Holders of licenses to operate
commercial nuclear plants under this
part must control public doses and dose
rates in unrestricted areas from normal
plant operations to meet the
requirements in 10 CFR part 20.
§ 53.270

Protection of plant workers.

Holders of licenses to operate
commercial nuclear plants under this
part must control occupational doses to
meet the requirements in 10 CFR part
20.
Subpart C—Design and Analysis
Requirements
§ 53.400 Design features for licensingbasis events.

(a) Design features must be provided
for each commercial nuclear plant such
that, when combined with
corresponding human actions and

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programmatic controls, the plant will
satisfy the safety criteria defined in
§§ 53.210 and 53.220, or more restrictive
alternative criteria adopted under
§ 53.470.
(b) Design features must ensure that
the safety functions identified in
§ 53.230 are fulfilled during licensingbasis events (LBEs).
§ 53.410 Functional design criteria for
design-basis accidents.

(a) Functional design criteria must be
defined for each design feature required
by § 53.400 and relied upon to
demonstrate compliance with the safety
criteria defined in § 53.210.
(b) Corresponding human actions and
programmatic controls must be
identified and implemented in
accordance with this and other subparts
to achieve and maintain the reliability
and capability of structures, systems,
and components (SSCs) relied upon to
satisfy the defined functional design
criteria and the safety criteria required
in § 53.210, and to maintain consistency
with analyses required by § 53.450(f).
§ 53.415 Protection against external
hazards.

Safety-related (SR) SSCs must be
protected against or must be designed to
withstand the effects of natural
phenomena (e.g., earthquakes,
tornadoes, hurricanes, floods, tsunami,
and seiches) and constructed hazards
(e.g., dams, transportation routes,
military and industrial facilities)
considering an event severity up to the
design-basis external hazard levels as
determined under § 53.510 without
losing the capability to perform the
safety functions identified under
§ 53.230. Specific requirements for
earthquake engineering are included in
§ 53.480.
§ 53.420 Functional design criteria for
licensing-basis events other than designbasis accidents.

(a) Functional design criteria must be
defined for each design feature required
by § 53.400 and relied upon to—
(1) Demonstrate compliance with the
safety criteria in § 53.220 or more
restrictive alternative criteria adopted
under § 53.470; and
(2) Demonstrate compliance with the
evaluation criteria in § 53.450(e) or more
restrictive alternative criteria adopted
under § 53.470.
(b) Corresponding human actions and
programmatic controls must be
identified and implemented in
accordance with this and other subparts
to achieve and maintain the reliability
and capability of SSCs relied upon to—

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(1) Satisfy the safety criteria in
§ 53.220 or more restrictive alternative
criteria adopted under § 53.470; and
(2) Satisfy the evaluation criteria in
§ 53.450(e) or more restrictive alternate
criteria adopted under § 53.470.
§ 53.425 Design features and functional
design criteria for normal operations.

(a) Design features must be provided
for each commercial nuclear plant to
support the Radiation Protection
Program required in § 53.850.
(b) Functional design criteria must be
defined for each design feature relied
upon to demonstrate compliance with
§ 53.850.
(c) Functional design criteria,
including design objectives for dose to
the maximally exposed member of the
public, must be defined for design
features to show that plant design
features and corresponding
programmatic controls, including
monitoring programs, control liquid,
gaseous, and solid wastes, as required
under part 20 of this chapter.1
1 A guide for keeping doses to the public
as low as is reasonably achievable is that the
estimated annual dose to the maximally
exposed member of the public does not
exceed 10 mrem total effective dose
equivalent. A design objective of maintaining
doses below 10 mrem/year should not be
construed as a radiation protection standard.

§ 53.430 Design features and functional
design criteria for protection of plant
workers.

(a) Design features must be provided
for each commercial nuclear plant such
that, when combined with
corresponding programmatic controls,
the requirements in § 53.270 can be met.
(b) Functional design criteria must be
defined for each design feature relied
upon to demonstrate compliance with
§ 53.270.

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§ 53.440

Design requirements.

(a)(1) Analysis, appropriate test
programs, prototype testing, operating
experience, or a combination thereof
must demonstrate that each design
feature required by § 53.400 meets the
defined functional design criteria
required by §§ 53.410 and 53.420. This
demonstration must consider
interdependent effects throughout the
commercial nuclear plant and the range
of conditions under which the design
features required by § 53.400 must
function throughout the plant’s lifetime.
(2) The design processes for SR and
non-safety-related but safety-significant
(NSRSS) SSCs under this part must
include administrative procedures for
evaluating operating, design, and
construction experience and for
considering applicable important

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industry experiences in the design of
those SSCs.
(b) The design features required by
§ 53.400 must, wherever applicable, be
designed using generally accepted
consensus codes and standards that
have been endorsed or otherwise found
acceptable by the U.S. Nuclear
Regulatory Commission (NRC).
(c) The materials used for each SR and
NSRSS SSC must be qualified for their
service conditions over the design life of
the SSC.
(d) Possible degradation mechanisms
related to aging, fatigue, chemical
interactions, operating temperatures,
effects of irradiation, and other
environmental factors that may affect
the performance of SR and NSRSS SSCs
must be evaluated and used to inform
the design and the development of
integrity assessment programs under
§ 53.870.
(e)(1) Safety-related and NSRSS SSCs
must be designed and located to
minimize, consistent with other safety
requirements in this part, the
probability and effect of fires and
explosions.
(2) Noncombustible and fire-resistant
materials must be used wherever
practical throughout the facility,
particularly in locations with SR and
NSRSS SSCs.
(3) Fire detection and fire suppression
systems of appropriate capacity and
capability must be provided and
designed to minimize the adverse effects
of fires on SR and NSRSS SSCs.
(4) Fire suppression systems must be
designed to ensure that their rupture or
inadvertent operation does not
significantly impair the ability of SR
and NSRSS SSCs to perform their safety
functions to satisfy § 53.230.
(f) Safety and security must be
considered together in the design
process such that, where possible,
security issues are effectively resolved
through design and engineered security
features.
(g) The reactor system and waste
stores for each commercial nuclear plant
must be capable of achieving and
maintaining a subcritical condition
during normal operations and following
any LBE identified in accordance with
§ 53.240.
(h) Each commercial nuclear plant
must have a capability to provide longterm cooling of the reactor fuel and
waste stores during normal operations
and following any LBE identified in
accordance with § 53.240.
(i) The design, analysis, staffing, and
programmatic controls for each
commercial nuclear plant must consider
the number of reactors, waste stores,
and other significant inventories of

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radioactive materials and the associated
operating configurations, common
systems, system interfaces, and system
interactions.
(j)(1) Design features must be
provided and related functional design
criteria defined such that, with limited
use of operator actions, one or more
physical barriers are maintained to limit
the release of radionuclides from reactor
systems, waste stores, or other
significant inventories of radioactive
materials assuming the impact of a
large, commercial aircraft.
(2) The functional design criteria for
those design features provided to
address the requirements in paragraph
(j)(1) of this section must be based on an
assessment of the impact of a large,
commercial aircraft used for long
distance flights in the United States,
with aviation fuel loading typically used
in such flights, and an impact speed and
angle of impact considering the ability
of both experienced and inexperienced
pilots to control large, commercial
aircraft at low altitude representative of
a commercial nuclear plant’s low
profile.1
1 Changes to the detailed parameters on
aircraft impact characteristics set forth in
guidance must be approved by the
Commission.

(k) Design features and related
functional design criteria must be
defined such that analyses demonstrate
a low risk of permanent injury to the
public due to the health effects of the
chemical hazards of licensed material.
(l) Measures must be taken during the
design of commercial nuclear plants to
minimize, to the extent practicable,
contamination of the facility and the
environment, facilitate eventual
decommissioning, and minimize, to the
extent practicable, the generation of
radioactive waste in accordance with
§ 20.1406 of this chapter.
(m)(1) Each commercial nuclear plant
must include criticality monitoring
capabilities meeting the requirements of
either § 70.24 of this chapter or
paragraph (m)(2) of this section.
(2) In lieu of maintaining a monitoring
system capable of detecting criticality as
described in § 70.24 of this chapter,
criticality accident requirements may be
satisfied by—
(i) Demonstrating the sub-criticality of
special nuclear material, except when it
is inside the reactor and the reactor is
being operated, by maintaining keffective below 0.95 at a 95 percent
probability, 95 percent confidence level,
under conditions that maximize
reactivity for the applicable storage and
handling configurations, and
(ii) Providing radiation monitors for
fuel storage and associated handling

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areas when fuel is present to detect
excessive radiation levels and to
support initiating appropriate safety
actions.
(3) While a spent fuel transportation
package approved under 10 CFR part 71
of this chapter or spent fuel storage cask
approved under 10 CFR part 72 is in the
special nuclear material handing or
storage area, the requirements in 10 CFR
parts 71 or 72, as applicable, and the
requirements of the certificate of
compliance for that package or cask, are
the applicable requirements for the fuel
within that package or cask.
(n)(1) The design of each commercial
nuclear plant must reflect state-of-theart human factors principles for safe and
reliable performance in all locations that
human activities are expected for
performing or supporting the continued
availability of plant safety or emergency
response functions.
(2) The design must provide for the
capabilities described in § 53.730(b) to
ensure the plant staff are able to monitor
plant conditions and respond to events.
(3) The means by which the design
and human actions together will achieve
the safety requirements of subpart B of
this part must be evaluated and used to
inform the design and the development
of the concept of operations required by
§ 53.730(c).
(4) A functional requirements analysis
and function allocation must be used to
ensure that plant design features
address how safety functions and
functional safety criteria are satisfied,
and how the safety functions will be
assigned to appropriate combinations of
human action, automation, active safety
features, passive safety features, or
inherent safety characteristics.

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§ 53.450

Analysis requirements.

(a) Requirement to have a
probabilistic risk assessment (PRA). A
PRA of each commercial nuclear plant
must be performed to identify potential
failures, susceptibility to internal and
external hazards, and other contributing
factors to event sequences that might
challenge the safety functions identified
in § 53.230 and to support
demonstrating that each commercial
nuclear plant meets the safety criteria of
§ 53.220, or more restrictive alternative
criteria adopted under § 53.470.
(b) Specific uses of analyses. The PRA
in combination with other generally
accepted approaches for systematically
evaluating engineered systems must be
used—
(1) In informing the selection of the
LBEs, as described in § 53.240, which
must be considered in the design to
determine compliance with the safety
criteria in subpart B of this part.

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(2) For informing the classification of
SSCs according to their safety
significance in accordance with § 53.460
and for identifying the environmental
conditions under which the SSCs and
operating staff must perform their safety
functions.
(3) In evaluating the adequacy of
defense-in-depth measures required in
accordance with § 53.250.
(4) To identify and assess all plant
operating states where there is the
potential for the uncontrolled release of
radioactive material to the environment.
(5) To identify and assess events that
challenge plant control and safety
systems whose failure could lead to the
uncontrolled release of radioactive
material to the environment. These
include internal events, such as human
errors and equipment failures, and
external events identified in accordance
with subpart D of this part.
(c) Maintenance and upgrade of
analyses. The PRA must be maintained
at least every 5 years until the
permanent cessation of operations
under § 53.1070 and upgraded in
conformance with generally accepted
methods, standards, and practices that
have been endorsed or otherwise found
acceptable by the NRC.
(d) Qualification of analytical codes.
The analytical codes used in modeling
plant behavior in analyses of licensingbasis events (including but not limited
to thermodynamics, reactor physics,
fuel performance, and mechanistic
source term codes) must be qualified for
the range of conditions for which they
are to be used.
(e) Analyses of licensing-basis events
other than design-basis accidents.
(1) Analyses must be performed for
LBEs other than design-basis accidents
(DBAs). These LBEs must be identified
using insights from a PRA in
combination with other generally
accepted approaches for systematically
evaluating engineered systems to
identify and analyze equipment failures
and human errors.
(2) The analysis of LBEs other than
DBAs must include definition of
evaluation criteria for each event or
specific categories of LBEs to determine
the acceptability of the plant response to
the challenges posed by internal and
external hazards to provide an
appropriate level of safety.
(3) The analyses of LBEs other than
DBAs must address event sequences
from initiation to a defined end state
and be used in combination with other
engineering analyses to demonstrate
that the functional design criteria
required by § 53.420 provide sufficient
barriers to the unplanned release of
radionuclides to satisfy the evaluation

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criteria defined for each LBE other than
DBAs, to satisfy the safety criteria
specified in accordance with § 53.220
and provide defense in depth as
required by § 53.250.
(4) The methodology used to identify,
categorize, and analyze LBEs must
include a means to identify event
sequences deemed significant for
controlling the risks posed to public
health and safety.
(f) Analysis of design-basis accidents.
(1) The analysis of LBEs required by
§ 53.240 must include analysis of DBAs
that address possible challenges to the
safety functions identified under
§ 53.230. The events selected as DBAs
must be those that, if not terminated,
have the potential for exceeding the
safety criteria in § 53.210.
(2) The DBAs selected must be
analyzed using deterministic methods
that address event sequences from
initiation to a safe stable end state and
assume only the SR SSCs identified
under § 53.460 and human actions
addressed by the requirements of
subpart F of this part are available to
perform the safety functions identified
in accordance with § 53.230.
(3) The analysis must conservatively
demonstrate compliance with the safety
criteria in § 53.210.
(g) Other required analyses. Analyses
must be performed to assess—
(1) Fire protection. Fire protection
measures to demonstrate, through
inclusion of fires in the analysis of LBEs
or by separate analyses, that a fire or
explosion in any plant area would not—
(i) Prevent equipment from fulfilling
the safety functions identified in
accordance with § 53.230, or
(ii) Challenge the safety criteria in
§§ 53.210 and 53.220.
(2) Aircraft impact. Measures
provided to protect against aircraft
impacts under § 53.440(j).
(3) Dose to members of the public.
Measures taken under § 53.425,
including estimating—
(i) The quantity of each of the
principal radionuclides expected to be
released annually to unrestricted areas
in liquid effluents produced during
normal reactor operations and the dose
to the maximally exposed member of
the public in unrestricted areas.
(ii) The quantities of each of the
principal radionuclides of the gases,
halides, and particulates expected to be
released annually to unrestricted areas
in gaseous effluents produced during
normal reactor operations and the dose
to the maximally exposed member of
the public in unrestricted areas.
(iii) The annual external radiation
dose in unrestricted areas and the
maximally exposed member of the

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public in unrestricted areas due to
direct radiation from contained
radiation sources from the commercial
nuclear plant during normal reactor
operations.
§ 53.460 Safety categorization and special
treatments.

(a) Structures, systems, and
components must be classified
according to their safety significance.
The SSC categories must include
‘‘Safety-Related,’’ ‘‘Non-Safety-Related
but Safety-Significant,’’ and ‘‘NonSafety-Significant,’’ as defined in
subpart A of this part.
(b) For SR and NSRSS SSCs, the
conditions under which they must
perform their safety function in § 53.230
must be identified. Special treatments
must be established in accordance with
this and other subparts to provide
confidence that the SSCs will perform
under the service conditions and with
reliability consistent with the analysis
performed under § 53.450 to
demonstrate meeting the safety criteria
in §§ 53.210 and 53.220, or more
restrictive alternative criteria adopted
under § 53.470.
(1) The special treatments for SR SSCs
must include meeting the applicable
quality assurance requirements from
appendix B of part 50 of this chapter.
(2) The special treatments for NSRSS
SSCs and special treatments for SR SSCs
beyond those required under (b)(1) of
this section may include meeting
selected quality assurance requirements
from appendix B of part 50 of this
chapter when such treatment is needed
to address performance requirements,
equipment reliability, or uncertainties.
(c) Human actions needed to prevent
or mitigate LBEs must be identified, be
able to be performed reliably under the
postulated environmental conditions,
and be addressed by programs
established in accordance with subpart
F of this part to provide confidence that
those actions will be performed as
assumed in the analysis performed in
accordance with § 53.450 to
demonstrate meeting the criteria in
§§ 53.210, 53.220, and 53.450(e), or
more restrictive alternative criteria
adopted under § 53.470.

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§ 53.470 Maintaining analytical safety
margins used to justify operational
flexibilities.

Where an applicant or licensee so
chooses, alternative criteria more
restrictive than those defined in
§§ 53.220 and 53.450(e) may be adopted
to support operational flexibilities. In
such cases, applicants and licensees
must ensure that the functional design
criteria of § 53.420, the analysis

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requirements of § 53.450(e), and
identification of special treatment of
SSCs and human actions under § 53.460
reflect and support the use of alternative
criteria to justify operational
flexibilities. Licensees must ensure that
measures taken to provide the analytical
margins supporting operational
flexibilities are incorporated into design
features and programmatic controls and
are maintained within programs
required in other subparts.
§ 53.480

Earthquake engineering.

(a) Effects of earthquakes. Structures,
systems, and components classified as
SR or NSRSS must be able to withstand
the effects of earthquakes,
commensurate with the safety
significance of the SSC, without loss of
capability to perform their role in
fulfilling the safety functions required
by § 53.230.
(b) Definitions. For the purpose of this
section—
Design-Basis Ground Motions
(DBGMs) are the vibratory ground
motions for which certain SSCs must be
designed to remain functional.
Operating basis earthquake (OBE)
ground motion is the vibratory ground
motion for which those features of the
commercial nuclear plant necessary for
continued operation without undue risk
to the health and safety of the public are
designed to remain functional. The OBE
ground motion is used in § 53.720.
Response spectrum is a plot of the
maximum responses (acceleration,
velocity, or displacement) of idealized
single-degree-of-freedom oscillators as a
function of the natural frequencies of
the oscillators for a given damping
value. The response spectrum is
calculated for a specified vibratory
motion input at the oscillators’
supports.
Surface deformation is the distortion
of geologic strata on or near the ground
surface that occurs because of tectonic
forces that result from earthquakes.
(c) Design considerations—(1) DesignBasis Ground Motions. (i) The DBGMs
must be derived from the Site Ground
Motion Response Spectra developed in
accordance with § 53.510(c), by taking
into consideration the functional design
criteria of SSCs in accordance with
§§ 53.410 and 53.420. The horizontal
component of the DBGM(s) in the freefield at the foundation level of the
structures must be an appropriate
response spectrum that is determined
based on the risk-significance of SSCs
and their safety functions. In view of the
limited data available on vibratory
ground motion of strong earthquakes, it
is acceptable that the design response
spectra be smoothed spectra.

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(ii) The commercial nuclear plant
must be designed so that, if the DBGMs
occur, the following SSCs remain
functional and within applicable stress,
strain, and deformation limits:
(A) Structures, systems, and
components for which functional design
criteria are established in accordance
with § 53.410 or § 53.420; and
(B) Structures, systems, and
components classified as SR or NSRSS
commensurate with safety significance
in accordance with § 53.460.
(iii) In addition to seismic loads,
applicable concurrent normal operating,
functional, and accident-induced loads
must be taken into account in the design
of the SR SSCs and, commensurate with
safety significance, NSRSS SSCs.
(iv) The design of the commercial
nuclear plant must take into account the
possible effects of seismic-induced
ground disruption, such as fissuring,
lateral spreads, differential settlement,
liquefaction, and landsliding, on the
facility foundations.
(v) The SSCs fulfilling the safety
functions required by § 53.230 must be
demonstrated through design, testing, or
qualification methods to be able to
fulfill those safety functions during and
after the vibratory ground motion
associated with the DBGMs.
(vi) The evaluation of SSCs required
by this section to show they are able to
function during and after earthquake
ground motion must take into account
soil-structure interaction effects and the
expected duration of vibratory motion.
It is permissible to design for strain
limits in excess of yield strain in some
of these SSCs during the DBGMs and
under the postulated concurrent loads,
provided the necessary safety functions
are maintained.
(2) OBE Ground Motion. The OBE
Ground Motion must be characterized
by response spectra. The value of the
OBE Ground Motion must be set to onethird or less of the DBGMs response
spectra.
(3) [Reserved]
(4) Required seismic instrumentation.
Suitable instrumentation must be
provided so that the seismic response of
commercial nuclear plant SR SSCs or
NSRSS SSCs can be evaluated promptly
after an earthquake.
(d) Surface deformation. (1) The
potential for surface deformation must
be taken into account in the design of
the commercial nuclear plant by
providing reasonable assurance that in
the event of deformation, SSCs
classified as SR or NSRSS in accordance
with § 53.460 will remain functional.
(2) In addition to surface deformation
induced loads, the design of SSCs must
take into account, commensurate with

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safety significance, seismic loads and
applicable concurrent functional and
accident-induced loads.
(3) The design provisions for surface
deformation must be based on its
postulated occurrence in any direction
and azimuth and under any part of the
commercial nuclear plant, unless
evidence indicates this assumption is
not appropriate, and must take into
account the estimated rate at which the
surface deformation may occur.
(e) Seismically induced floods and
water waves and other design
conditions. Seismically induced floods
and water waves from either locally or
distantly generated seismic activity and
other design conditions determined
pursuant to subpart D of this part must
be taken into account in the design of
the commercial nuclear plant so as to
prevent undue risk to the health and
safety of the public.
(f) Analysis. The analyses required by
§ 53.450 must address seismic hazards
and related SSC responses in
determining that the safety criteria
defined in § 53.220 will be met.
(g) Design criteria, human actions,
and programmatic controls. Functional
design criteria, human actions, and
programmatic controls needed to
address seismic events must be
identified and implemented in
accordance with this and other subparts
to achieve and maintain the
performance of SSCs relied upon to
satisfy the safety criteria in § 53.220 and
to maintain consistency with analyses
required by § 53.450 when accounting
for the site-specific frequencies and
magnitudes of earthquakes for a
commercial nuclear plant.
Subpart D—Siting Requirements

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§ 53.500 General siting and siting
assessment.

(a) The siting of each commercial
nuclear plant must be supported by
assessments of proposed sites such that
the design, including design features
and programmatic controls
corresponding to the site characteristics,
satisfies the safety criteria defined in
§§ 53.210 and 53.220 or more restrictive
alternative criteria adopted under
§ 53.470. The siting assessment must
ensure that site characteristics that
might contribute to the initiation,
progression, or consequences of
licensing-basis events (LBEs) analyzed
under §§ 53.450 and 53.480 are
identified and mitigated by design
features or programmatic controls. The
siting assessment must take into
consideration the potential adverse
impacts that a commercial nuclear plant

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may have on nearby populations as a
result of normal operations or LBEs.
(b) Activities performed to identify
site characteristics or otherwise needed
to determine site-specific contributors to
functional design criteria or analysis
assumptions under subpart C of this
part must satisfy the applicable special
treatment requirements of § 53.460,
including, where applicable, the quality
assurance requirements from appendix
B of part 50 of this chapter.
§ 53.510

External hazards.

(a) General external hazard
requirements. The design-basis external
hazard level for the relevant external
hazards for a site must be identified and
characterized based on site-specific
assessments of natural and constructed
hazards with the potential to adversely
affect plant functions. The external
hazard frequencies and magnitudes
determined from the site-specific
assessments must take into account
uncertainties and variabilities in data,
models, and methods relied on to
characterize the external hazards.
(b) Definitions. For the purpose of this
section, the following terms mean:
Geological siting factors are geological
and seismic factors that may affect the
design and operation of the proposed
commercial nuclear plant.
Ground Motion Response Spectra
(GMRS) are the site-specific GMRS
resulting from the geologic
investigations and evaluations of the
site vicinity and region and used to
determine design-basis ground motions
for structures, systems, and components
under § 53.480.
Probabilistic seismic hazard analysis
is an analytical methodology that
incorporates uncertainty into estimates
of an annual frequency of exceedance
for a certain ground motion parameter
(e.g., peak ground acceleration, peak
ground velocity, response spectral
values) at a site.
(c) Geological investigations. The
GMRS for the site must be determined
based on the results of investigations of
the geological, seismological, and
engineering characteristics of the site
and its environs and must be
characterized by both horizontal and
vertical free-field GMRS at the free
ground surface. The size of the region to
be investigated and the type of data
pertinent to the investigations must be
determined based on the nature of the
region surrounding the site. Data on
vibratory ground motion, earthquake
recurrence rates, fault geometry and slip
rates, and site subsurface material
properties must be obtained by
reviewing pertinent literature and
carrying out field investigations.

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Uncertainties are inherent in the
parameters and models used to estimate
the GMRS for the site. The site
assessment must reflect these
uncertainties through an appropriate
analysis, such as a probabilistic seismic
hazard analysis.
(d) Geologic and seismic siting
factors. The geologic and seismic siting
factors considered for design under
§§ 53.415 and 53.480 must include, but
are not limited to, determination of the
potential for surface tectonic and
nontectonic deformations, the size and
character of seismically induced floods
and water waves that could affect a site
from either locally or distantly
generated seismic activity, soil and rock
stability, liquefaction potential, and
natural and artificial slope stability.
§ 53.520

Site characteristics.

Site characteristics that might
contribute to the initiation, progression,
or consequences of LBEs analyzed
under § 53.450 must be identified,
assessed, and considered in the design
and analyses required by subpart C of
this part.
§ 53.530 Population-related
considerations.

Every site must have an exclusion
area, a low-population zone, and a
population center distance as defined in
§ 53.020.
(a) The offsite radiological
consequences estimated by the analyses
required by § 53.450(f) must be used to
confirm that—
(1) An individual located at any point
on the boundary of the exclusion area
for any 2-hour period following onset of
the postulated fission product release
would not receive a radiation dose in
excess of 25 rem (250 millisieverts) total
effective dose equivalent.
(2) An individual located at any point
on the outer boundary of the lowpopulation zone who is exposed to the
radioactive cloud resulting from the
postulated fission product release
(during the entire period of its passage)
would not receive a radiation dose in
excess of 25 rem (250 millisieverts) total
effective dose equivalent.
(b) The population center distance
must be at least one and one-third times
the distance from the reactor to the
outer boundary of the low-population
zone. The boundary of the population
center must be determined upon
consideration of population
distribution. Political boundaries are not
controlling in the calculation of
population center distance.
(c) Reactor sites should be located
away from very densely populated
centers. Areas of low-population density

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are, generally, preferred. However, in
determining the acceptability of a
particular site located away from a very
densely populated center but not in an
area of low-population density,
consideration will be given to safety,
environmental, economic, or other
factors, which may result in the site
being found acceptable.
§ 53.540

Siting interfaces.

Site characteristics must be addressed
by the design features, programmatic
controls, and supporting analyses used
to demonstrate that the safety criteria in
§§ 53.210 and 53.220 are met for each
commercial nuclear plant. Site
characteristics must be such that
adequate emergency plans and security
plans can be developed and maintained.
Subpart E—Construction and
Manufacturing Requirements
§ 53.600 Construction and
manufacturing—scope and purpose.

This subpart applies to those
construction and manufacturing
activities authorized by a construction
permit (CP), combined license (COL),
manufacturing license (ML), or limited
work authorization (LWA) issued under
this part.

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§ 53.605 Reporting of defects and
noncompliance.

Each CP and ML issued under this
part is subject to the terms and
conditions in this section, and each COL
issued under this part is subject to the
terms and conditions in this section
until the date that the Commission
makes the finding under § 53.1452(g).
(a) Definitions. The definitions in
§ 21.3 of this chapter apply to this
section.
(b) Posting requirements. (1) Each
individual, partnership, corporation,
dedicating entity, or other entity subject
to the regulations in this section must
post current copies of this section and
the regulations in 10 CFR part 21;
section 206 of the Energy
Reorganization Act of 1974, as
amended; and procedures adopted
under these regulations. These
documents must be posted in a
conspicuous position on any premises
within the United States where the
activities subject to the license are
conducted.
(2) If posting of these regulations or
the procedures adopted under them is
not practical, the licensee may, in
addition to posting section 206 of the
Energy Reorganization Act of 1974, as
amended, post a notice that describes
the regulations/procedures, including
the name of the individual to whom

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reports may be made, and states where
they may be examined.
(c) Procedures. The holder of a CP,
COL, or ML subject to this section must
adopt appropriate procedures to—
(1) Evaluate deviations and failures to
comply to identify defects and failures
to comply associated with substantial
safety hazards as soon as practicable,
and, except as provided in paragraph
(c)(2) of this section, in all cases within
60 days of discovery, to identify a
reportable defect or failure to comply
that could create a substantial safety
hazard, were it to remain uncorrected.
(2) Ensure that if an evaluation of an
identified deviation or failure to comply
potentially associated with a substantial
safety hazard cannot be completed
within 60 days from the discovery of the
deviation or failure to comply, an
interim report is prepared and
submitted to the Commission through a
director or responsible officer, or
designated person as discussed in
paragraph (d)(5) of this section. The
interim report should describe the
deviation or failure to comply that is
being evaluated and should also state
when the evaluation will be completed.
This interim report must be submitted
in writing within 60 days of discovery
of the deviation or failure to comply.
(3) Ensure that a director or
responsible officer of the holder of a CP,
COL, or ML subject to this section is
informed as soon as practicable, and, in
all cases, within the 5 working days
after completion of the evaluation
described in paragraph (c)(1) or (c)(2) of
this section, if the construction or
manufacture of a facility or activity, or
a basic component supplied for such a
facility or activity—
(i) Fails to comply with the Atomic
Energy Act of 1954, as amended, or any
applicable regulation, order, or license
of the Commission relating to a
substantial safety hazard;
(ii) Contains a defect; or
(iii) Underwent any significant
breakdown in any portion of the quality
assurance program (QAP) conducted
under the requirements of appendix B to
part 50 of this chapter that could have
produced a defect in a basic component.
These breakdowns in the QAP are
reportable whether or not the
breakdown actually resulted in a defect
in a design approved and released for
construction, installation, or
manufacture.
(d) Reporting defects and
noncompliance. (1) The holder of a CP,
COL, or ML subject to this section that
obtains information reasonably
indicating that the facility or
manufactured reactors fails to comply
with the Atomic Energy Act of 1954, as

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amended, or any applicable regulation,
order, or license of the Commission
relating to a substantial safety hazard
must notify the Commission of the
failure to comply through a director,
responsible officer, or designated person
as discussed in paragraph (d)(5) of this
section.
(2) The holder of a CP, COL, or ML
subject to this section that obtains
information reasonably indicating the
existence of any defect found in the
construction or manufacture, or any
defect found in the final design of a
facility as approved and released for
construction or manufacture, must
notify the Commission of the defect
through a director, responsible officer,
or designated person as discussed in
paragraph (d)(5) of this section.
(3) The holder of a CP, COL, or ML
subject to this part, who obtains
information reasonably indicating that
the QAP has undergone any significant
breakdown discussed in paragraph
(c)(3)(iii) of this section must notify the
Commission of the breakdown in the
QAP through a director, responsible
officer, or designated person as
discussed in paragraph (d)(5) of this
section.
(4) When acting as a dedicating entity,
the holder of a CP, COL, or ML subject
to this section is responsible for
identifying and evaluating deviations;
reporting defects and failures to comply
associated with substantial safety
hazards for dedicated items; and
maintaining auditable records for the
dedication process.
(5) The notification requirements of
this paragraph apply to all defects and
failures to comply associated with a
substantial safety hazard regardless of
whether extensive evaluation, redesign,
or repair is required to conform to the
criteria and bases stated in the Safety
Analysis Report, CP, COL, or ML.
Evaluation of potential defects and
failures to comply and reporting of
defects and failures to comply under
this section satisfies the CP holder’s,
COL holder’s, and ML holder’s
evaluation and notification obligations
under 10 CFR part 21, and satisfies the
responsibility of individual directors or
responsible officers or holders of a CP,
COL, or ML subject to this section to
report defects, and failures to comply
associated with substantial safety
hazards under section 206 of the Energy
Reorganization Act of 1974, as
amended. The director or responsible
officer may authorize an individual to
provide the notification required by this
section. However, this does not relieve
the director or responsible officer of his
or her responsibility under this section.

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(e) Notification—timing and where
sent. The notification required by
paragraph (d) of this section must
consist of—
(1) Initial notification by telephone,
facsimile, or email identified in
appendix A to 10 CFR part 73 to the
U.S. Nuclear Regulatory Commission
(NRC) Operations Center within 2 days
following receipt of information by the
director or responsible corporate officer
under paragraph (c)(3) of this section,
on the identification of a defect or a
failure to comply. If the CP, COL, or ML
holder elects to use facsimile,
verification that the facsimile has been
received should be made by calling the
NRC Operations Center. This paragraph
does not apply to interim reports
described in paragraph (c)(2) of this
section.
(2) Written notification submitted to
the NRC Document Control Desk by an
appropriate method listed in § 53.040,
with a copy to the appropriate NRC
Regional Administrator at the address
specified in appendix D to 10 CFR part
20 and a copy to the appropriate NRC
resident inspector, if applicable, within
30 days following receipt of information
by the director or responsible corporate
officer under paragraph (c)(3) of this
section, on the identification of a defect
or failure to comply.
(f) Content of notification. The written
notification required by paragraph (e)(2)
of this section must clearly indicate that
the written notification is being
submitted under this section and
include the following information, to
the extent known.
(1) Name and address of the
individual or individuals informing the
Commission.
(2) Identification of the facility, the
activity, or the basic component
supplied for the facility or the activity
within the United States which contains
a defect or fails to comply.
(3) Identification of the firm
constructing or manufacturing the
facility or supplying the basic
component which fails to comply or
contains a defect.
(4) Nature of the defect or failure to
comply and the safety hazard which is
created or could be created by the defect
or failure to comply.
(5) The date on which the information
of a defect or failure to comply was
obtained.
(6) In the case of a basic component
that contains a defect or failure to
comply, the number and location of
these components in use at the facility
subject to the regulations in this part.
(7) In the case of a completed reactor
manufactured under this part, the

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entities to which the reactor was
supplied.
(8) The corrective action which has
been, is being, or will be taken; the
name of the individual or organization
responsible for the action; and the
length of time that has been or will be
taken to complete the action.
(9) Any advice related to the defect or
failure to comply about the facility,
activity, or basic component that has
been, is being, or will be given to other
entities.
(g) Procurement documents. Each
holder of a CP, COL, or ML subject to
this section must ensure that each
procurement document for a facility or
a basic component specifies the
provisions of 10 CFR part 21 or this
section that apply, as applicable.
(h) Coordination with 10 CFR part 21.
The requirements of this section are
satisfied when the defect or failure to
comply associated with a substantial
safety hazard has been previously
reported under 10 CFR part 21, under
§ 73.1205 of this chapter, under this
section, or under § 53.1640.
(i) Records retention. The holder of a
CP, COL, or ML subject to this section
must prepare and maintain records
necessary to accomplish the purposes of
this section, specifically—
(1) Retain procurement documents,
which define the requirements that
facilities or basic components must
satisfy in order to be considered
acceptable, for the lifetime of the facility
or basic component.
(2) Retain records of evaluations of all
deviations and failures to comply under
paragraph (c)(1) of this section for the
longest of—
(i) Ten years from the date of the
evaluation;
(ii) Five years from the date that an
early site permit is referenced in an
application for a COL; or
(iii) Five years from the date of
delivery of a manufactured reactor.
(3) Retain records of all interim
reports to the Commission made under
paragraph (c)(2) of this section, or
notifications to the Commission made
under paragraph (d) of this section for
the minimum time periods stated in
paragraph (i)(2) of this section;
(4) Suppliers of basic components
must retain records of—
(i) All notifications sent to affected
licensees or purchasers under paragraph
(d)(4) of this section for a minimum of
10 years following the date of the
notification;
(ii) The facilities or other purchasers
to whom the basic components or
associated services were supplied for a
minimum of 15 years from the delivery

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of the basic component or associated
services.
(5) Maintaining reports in accordance
with this section satisfies the
recordkeeping obligations under 10 CFR
part 21 of the entities, including
directors or responsible officers thereof,
subject to this section.
§ 53.610

Construction.

(a) Management and control.
Licensees must ensure that the
following plans, programs, and
organizational units are developed and
implemented to manage and control the
construction activities:
(1) Programs to ensure that the
construction of a commercial nuclear
plant supports the eventual compliance
with the design and analysis
requirements in subpart C of this part.
(2) An organization, headed by
qualified personnel, responsible for
managing, controlling, and evaluating
the adequacy of the construction
activities.
(3) Procedures describing the
qualifications for personnel in key
positions in the licensee’s management
and control organization and the
organizational responsibilities,
authority, and interfaces with other
parts of the licensee’s organization.
(4) Procedures to evaluate the
applicability of other national and
international construction experience to
the planned and ongoing construction
activities and to ensure the applicable
experience will be provided to those
constructing the plant.
(5) A fitness-for-duty program, under
10 CFR part 26.
(6)(i) A QAP meeting the
requirements of appendix B of part 50
of this chapter as required by
§ 53.460(b).
(ii) Appropriate programmatic
controls to provide special treatment for
non-safety-related but safety-significant
structures, systems, and components
(SSCs).
(7) A radiation protection program, in
accordance with 10 CFR part 20, that
includes measures for monitoring the
dose to individuals working with
radioactive materials brought onto the
site, as applicable.
(8) An information security program
in accordance with §§ 73.21, 73.22, and
73.23 of this chapter, as applicable.
(b) Construction activities. No person
may begin the construction of a
commercial nuclear plant on a site on
which the facility is to be operated
under this part until that person has
been issued either a CP or COL, an early
site permit authorizing activities under
§ 53.1130, or an LWA under this part.
(1) Licensees must satisfy the
following requirements:

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(i) As appropriate, considering the
types and quantities of radioactive
materials being brought onto the site—
(A) The licensee must maintain and
follow a special nuclear material (SNM)
material control and accounting
program, a measurement control
program, and other material control
procedures that include corresponding
record management requirements as
required by the provisions of § 70.32 of
this chapter. Prior to initial receipt of
SNM onsite, the licensee must
implement an SNM material control and
accounting program in accordance with
10 CFR part 74.
(B) Procedures must be in place to
receive, possess, use, and store source,
byproduct, and SNM in accordance with
applicable portions of 10 CFR parts 30,
40, and 70.
(C) A plant staff training program
associated with the receipt of
radioactive material must be approved
and implemented prior to initial receipt
of byproduct, source or SNM (excluding
exempt quantities as described in
§ 30.18 of this chapter).
(ii) For construction of a commercial
nuclear plant involving multiple reactor
units, plans and procedures must be in
place to prevent or mitigate potential
hazards to the SSCs of operating units
resulting from construction activities,
including the managerial and
administrative controls to be used to
provide assurance that the limiting
conditions for operation of the operating
units are not exceeded as a result of
construction activities.
(iii) Procedures must be in place prior
to the start of construction activities that
describe how construction will be
controlled so as not to impact other
features important to the design, such as
dewatering, slope stability, backfill,
compaction, and seepage.
(iv) For LWA holders, a plan must be
developed for redress of activities
performed under the LWA should one
of the following situations arise:
(A) LWA work activities are
terminated by the holder of the LWA;
(B) The LWA is revoked by the NRC;
or
(C) The Commission denies the
associated CP or COL application.
(2)(i) Onsite fresh fuel must be
protected and stored in compliance with
§ 73.67 of this chapter.
(ii) Before initial fuel load into the
reactor (or, for a fueled manufactured
reactor, before initiating the physical
removal of any one of the independent
physical mechanisms to prevent
criticality required under
§ 53.620(d)(1)), a cybersecurity program
that meets the requirements of §§ 73.54
or 73.110 of this chapter, a physical

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security program that meets the
requirements of §§ 73.55 or 73.100 of
this chapter, and an access
authorization program that meets the
requirements of §§ 73.56 or 73.120 of
this chapter must be established, as
applicable.
(iii) Fire protection measures must be
implemented for work and storage areas
(including adjacent fire areas that could
affect the work or storage area) before
initial receipt of byproduct, source, or
non-fuel SNM (excluding exempt
quantities as described in § 30.18 of this
chapter). The fire protection measures
for areas associated with new fuel
(including all fuel handling, fuel
storage, and adjacent fire areas that
could affect the new fuel) must be
implemented before receipt of fuel.
Prior to the receipt of fuel, a formal
letter of agreement must be in place
with the local fire department
specifying the nature of arrangements in
support of the fire protection program.
(c) Inspection and acceptance. (1) The
licensee must have a process for
accepting individual or groups of SSCs
upon completion of construction and
protecting them from damage or
tampering as other construction
activities continue.
(2) The post construction acceptance
process must address the inspections,
tests, analyses, and acceptance criteria
specified in the COL under § 53.1440 or
the equivalent verifications needed to
support the issuance of an operating
license under § 53.1387.
§ 53.620

Manufacturing.

(a) Management and control. Holders
of MLs must ensure that the following
plans, programs, and organizational
units are developed and implemented to
manage and control the manufacturing
activities within the scope of the ML:
(1) Programs to ensure that the
manufacturing of a manufactured
reactor or portions of a manufactured
reactor complies with the design and
analysis requirements in subpart C of
this part. The entity with design
authority for the manufactured reactor
covered by the ML must be identified in
the license.
(2) An organizational and
management structure responsible for
managing, controlling, and evaluating
the adequacy of the reactor design and
manufacturing activities.
(3) Procedures describing the
qualifications for personnel in key
positions in the licensee’s management
and control organization and the
organizational responsibilities,
authority, and interfaces with other
parts of the licensee’s organization.

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(4) A program to evaluate the
applicability of other national and
international design and manufacturing
experience to the planned and ongoing
manufacturing activities.
(5) A fitness-for-duty program, in
accordance with 10 CFR part 26.
(6)(i) A QAP meeting the
requirements of appendix B to part 50
of this chapter, to be applied to the
design, fabrication, construction, and
testing of the SSCs of the manufactured
reactor.
(ii) Appropriate programmatic
controls to provide special treatment
measures for non-safety-related but
safety-significant SSCs.
(7) A radiation protection program, in
accordance with 10 CFR part 20, that
includes measures for monitoring the
dose to individuals if the manufacturing
activities include working with
radioactive materials.
(8) An information security program
in accordance with §§ 73.21, 73.22 and
73.23 of this chapter, as applicable.
(b) Manufacturing activities. Holders
of MLs must satisfy the following
requirements:
(1) The manufacturing process must
be conducted within facilities for which
the ML holder has the authority to
establish controls on any activity that
might affect manufacturing. The
licensee must establish access controls
to the portions of each facility involved
in the manufacturing processes
governed by the ML.
(2) Manufacturing processes must be
performed in accordance with the ML
and the referenced codes and standards
that have been endorsed or otherwise
found acceptable by the NRC.
(3) A post-manufacturing inspection
and acceptance process must be
established and implemented before
transporting a manufactured reactor or
portions of a manufactured reactor for
installation at a commercial nuclear
plant. The process must consider the
results of inspections, tests, and
analyses that have been performed and
the acceptance criteria that are
necessary and sufficient to conclude
that manufacturing activities have been
completed in accordance with the ML.
(c) Control of radioactive materials.
As appropriate considering the types
and quantities of radioactive materials
being brought into the manufacturing
facility—
(1) Procedures must be in place to
receive, transfer, possess, and use
source, byproduct, and SNM in
accordance with the applicable portions
of 10 CFR parts 30, 40 and 70.
(2) A fire protection program must be
established and implemented before the
initial receipt of byproduct, source, or

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non-fuel SNM (excluding exempt
quantities as described in § 30.18 of this
chapter).
(3) An emergency plan appropriate for
responding to the facility-specific
hazards of an accidental release of
radioactive material and to limit the
health effects of the associated chemical
hazards of licensed material must be
approved and implemented prior to the
receipt of byproduct, source, or SNM
(excluding exempt quantities as
described in § 30.18 of this chapter).
(4) A plant staff training program
associated with the receipt of
radioactive material must be approved
and implemented before initial receipt
of byproduct, source, or SNM
(excluding exempt quantities as
described in § 30.18 of this chapter).
(5) Security requirements must be
implemented for the protection of SNM
based on the type, enrichment, and
quantity in accordance with 10 CFR part
73, as applicable, and for the protection
of Category 1 and Category 2 quantities
of radioactive material in accordance
with 10 CFR part 37, as applicable.
(d) Fuel loading. (1)(i) An ML may
authorize possession of a manufactured
reactor into which the licensee has
loaded fresh (unirradiated) fuel
pursuant to a license issued under part
70 of this chapter only if the
manufactured reactor is configured
during its loading, storage, and transport
with at least two independent physical
mechanisms in place, each of which is
sufficient to prevent criticality assuming
optimum neutron moderation and
neutron reflection conditions.
(ii) The ML applicant may file a
separate, subsequent application for the
10 CFR part 70 license or combine the
application for the 10 CFR part 70
license with the application for an ML.
(iii) The Commission has determined
that any such fueled manufactured
reactor in which the independent
physical mechanisms to prevent
criticality have been installed is not in
operation.
(iv) Upon installation of the fueled
manufactured reactor in its place of
operation and a Commission finding
that the acceptance criteria in the COL
that authorized reactor construction are
met under § 53.1452(g), the independent
physical mechanisms to prevent
criticality may be removed. Upon
initiating the physical removal of any
one of the independent physical
mechanisms to prevent criticality, the
fueled manufactured reactor has
commenced operation.
(2) Holders of 10 CFR part 70 licenses
authorizing the possession and loading
of fresh fuel into manufactured reactors
must comply with the requirements of

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10 CFR part 70 for the facilities and
activities related to the storage,
movement, and loading of fresh fuel in
the manufactured reactor. Holders of
these 10 CFR part 70 licenses must
comply with the requirements of
Subpart H to 10 CFR part 70, regardless
of whether their proposed activities
meet the applicability criteria found in
10 CFR 70.60. Procedures, equipment,
and personnel required by the 10 CFR
part 70 license, must be in place before
the receipt of SNM at the manufacturing
facility.
(i) Before the receipt of SNM, the
licensee must have security programs in
place that meet the performance
objectives of 10 CFR 73.67, with the
following additions and exceptions:
(A) A physical security plan
describing the physical security
program must be maintained and a
cybersecurity program must be
established for the possession and
loading of fresh fuel into a
manufactured reactor authorized by a 10
CFR part 70 license, regardless of fuel
type, enrichment, and quantity.
(B) The physical security program
must be designed to prevent unintended
and uncontrolled criticality events.
(C) The cybersecurity program must
provide reasonable assurance that a
cyberattack would not adversely impact
the functions performed by digital assets
used by the licensee for implementing
the physical security requirements of
this section, or the radiation monitoring
and criticality requirements in this
section or in 10 CFR part 70.
(D) All holders of a part 70 license
that authorizes loading of fresh fuel into
a manufactured reactor must perform
the screening required in § 73.67(d)(4) of
this chapter to confirm the identity,
trustworthiness, and reliability of
individuals prior to granting unescorted
access to special nuclear material; these
determinations must be documented.
(ii) [Reserved]
(3) The loading or unloading of fresh
fuel into or from a manufactured reactor
and any changes to the configuration of
reactivity control and prevention
systems for the fueled manufactured
reactor must be performed by a certified
fuel handler meeting the requirements
in subpart F of this part.
(e) Transportation. (1) A holder of an
ML may not transport or allow to be
removed from the places of manufacture
the manufactured reactor or portions
thereof as defined in the ML except for
transport to a site for which the
Commission has issued a COL that
references the subject ML.
(2) A holder of an ML must include
in any contract governing the transport
of a manufactured reactor or portions

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thereof as defined in the ML from the
places of manufacture to any other
location, a provision requiring that the
person transporting the manufactured
reactor comply with all shipping
requirements in applicable NRC
regulations, certificates of compliance,
and NRC-issued licenses.
(3) Procedures governing the
preparation of the manufactured reactor
or portions thereof as defined in the ML
for transport and the conduct of the
transport must be issued prior to
transport. The procedures must
implement the protective measures and
restrictions described in NRC
regulations and NRC-issued licenses to
protect the reactor from potential
conditions that would adversely affect
the safe operation of a commercial
nuclear plant.
(4) For a manufactured reactor that is
to be loaded with fresh fuel before
transport to the place of operation, the
ML must specify that transportation will
be in accordance with parts 71 and 73
of this chapter.
(f) Acceptance and installation at the
site for which the Commission has
issued a COL that references the subject
ML. (1) Installation at the site for which
the Commission has issued a COL that
references the subject ML must follow
the regulations in § 53.610.
(2) Upon arrival at the site, the
manufactured reactor or portions of a
manufactured reactor may not be
installed in its place of operation unless
the COL holder performs inspections
sufficient to verify the reactor is in
compliance with the ML and has not
been damaged in transit. The COL
holder must perform these inspections
in accordance with documented
procedures subject to quality assurance
measures commensurate with their
importance to safety. In addition,
inspections must confirm that the
interface requirements between the
manufactured reactor or portions of a
manufactured reactor and the remaining
portions of the commercial nuclear
plant are met.
Subpart F—Requirements for
Operation
§ 53.700

Operational objectives.

(a) Each holder of an operating license
(OL) or combined license (COL) under
this part must develop, implement, and
maintain controls for plant structures,
systems, and components (SSCs),
responsibilities of plant personnel, and
plant programs during the operating life
of each commercial nuclear plant such
that the requirements defined in subpart
B are satisfied. More specifically:

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(1) Each holder of an OL or COL
under this part must maintain the
capabilities, availability, and reliability
of plant SSCs to ensure that the safety
functions identified in § 53.230 will be
performed if called upon during
licensing-basis events (LBEs).
(2) Each holder of an OL or COL
under this part must ensure that plant
personnel have adequate knowledge and
skills to perform their assigned duties
that support the performance of the
safety functions identified in § 53.230.
(3) Each holder of an OL or COL
under this part must implement plant
programs sufficient to ensure that the
safety functions identified in § 53.230
will be performed if called upon during
normal operations and LBEs.
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§ 53.710 Maintaining capabilities and
availability of structures, systems, and
components.

Controls must be provided for each
commercial nuclear plant licensed
under this part such that the
capabilities, availability, and reliability
of plant SSCs, when combined with
corresponding programmatic controls
and human actions, provide that the
safety criteria defined in §§ 53.210 and
53.220 will be met.
(a) Technical specifications must be
developed, implemented, and
maintained that define conditions or
limitations on plant operations that are
necessary to ensure that safety-related
(SR) SSCs can fulfill the safety functions
identified under § 53.230 and support
meeting the safety criteria of § 53.210.
The technical specifications must
describe the following requirements:
(1) Limits on the inventory of
radioactive materials within the reactor
system and supporting systems with the
potential, individually or collectively, to
cause a release exceeding the safety
criteria in § 53.210 as a result of a
design-basis accident analyzed in
accordance with § 53.450(f).
(2) Operating limits for the facility
that if exceeded could lead to a failure
to perform a required safety function
necessary to demonstrate compliance
with the safety criteria in § 53.210.
(3) For each SSC classified as SR in
accordance with § 53.460, technical
specifications must define—
(i) Limiting conditions for operation.
Limiting conditions for operation are
the lowest functional capability or
performance levels of SR SSCs required
to ensure that the design-basis accidents
analyzed in accordance with § 53.450(f)
satisfy the safety criteria of § 53.210.
When a limiting condition for operation
is not met, the licensee must shut down
the plant or follow any remedial action

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permitted by the technical
specifications until the condition can be
met.
(ii) Surveillance requirements.
Surveillance requirements are
requirements relating to test, calibration,
or inspection to assure that the
necessary quality of systems and
components is maintained and that the
limiting conditions for operation will be
met.
(4) Design elements to be included are
those elements of the plant such as
materials of construction and geometric
arrangements, which, if altered or
modified, would have a significant
effect on safety and are not covered in
categories described in paragraphs (a)(1)
through (3) of this section.
(5) Administrative controls are the
provisions relating to organization and
management, procedures,
recordkeeping, review and audit, and
reporting necessary to assure operation
of the plant in a safe manner. Each
licensee must submit any reports to the
Commission pursuant to approved
technical specifications under § 53.040.
(b) Controls on plant operations,
including availability controls, must be
developed and implemented to ensure
that the configurations and special
treatments for SR SSCs and non-safetyrelated but safety-significant (NSRSS)
SSCs provide the capabilities,
availability, and reliability required to
demonstrate compliance with the
criteria of §§ 53.220 and 53.450(e).1 The
controls must—
1 The

comprehensive risk metrics and
related risk performance objectives
established under § 53.220 involve assessing
and averaging the risks over a defined period
(e.g., plant year) and do not constitute a realtime requirement that must be continuously
demonstrated by the licensee.

(1)(i) Identify who within the
commercial nuclear plant has authority
to make configuration changes;
(ii) Establish processes to make
configuration changes to NSRSS SSCs;
and
(iii) Establish processes to ensure that
all organizations of the commercial
nuclear plant affected by the
configuration changes are formally
notified and approve of the change.
(2) Describe how the special
treatments for each NSRSS SSC and
special treatments for SR SSCs beyond
those under paragraph (a) of this section
will be established and maintained over
the operating life of the commercial
nuclear plant.
§ 53.715 Maintenance, repair, and
inspection programs.

(a) A program to control maintenance
activities and monitor the performance

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or condition of SR and NSRSS SSCs
must be developed, implemented, and
maintained.
(b) Whenever a licensee determines
through activities related to
maintenance, repair, and inspection of
SSCs, the activities under § 53.710, or
otherwise that the performance or
condition of an SR or NSRSS SSC does
not demonstrate compliance with
established special treatments or
performance goals related to
capabilities, availability, or reliability,
the licensee must take appropriate
corrective action.
(c) Performance and condition
monitoring activities and associated
goals and preventive maintenance
activities must be evaluated at least
every 24 months. The evaluations must
take into account, where practical,
industry-wide operating experience.
Adjustments must be made where
necessary to ensure that the objective of
preventing failures of SSCs through
maintenance is appropriately balanced
against the objective of minimizing
unavailability of SSCs due to
monitoring or preventive maintenance.
(d) Before performing maintenance
activities (including but not limited to
surveillance, post-maintenance testing,
and corrective and preventive
maintenance), the licensee must assess
and manage the increase in risk that
may result from the proposed
maintenance activities.
§ 53.720

Response to seismic events.

If vibratory ground motion exceeding
that of the operating basis earthquake
ground motion or significant plant
damage due to vibratory ground motion
occurs, the licensee must shut down the
commercial nuclear plant. If structures,
systems, or components necessary for
the safe shutdown of the commercial
nuclear plant are not available after the
occurrence of this vibratory ground
motion, the licensee must consult with
the Commission and must propose a
plan for the timely, safe shutdown of the
commercial nuclear plant. Prior to
resuming operations, the licensee must
demonstrate to the Commission that
those features necessary for continued
operation without undue risk to the
health and safety of the public or
necessary to maintain the licensing
basis of the commercial nuclear plant
were either not functionally damaged or
have been repaired.
§ 53.725 General staffing, training,
personnel qualifications, and human factors
requirements.

(a) Two classes of commercial nuclear
plants. Commercial nuclear plants
licensed under this part are either of the

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class, based upon the similarity of
operating and technical characteristics
of the plants in the class, of self-reliantmitigation facilities or of interactiondependent-mitigation facilities. A
commercial nuclear plant is a selfreliant-mitigation facility if the U.S.
Nuclear Regulatory Commission (NRC)
determined as part of its approval of the
OL or COL for that plant that its design
demonstrates compliance with the
criteria of § 53.800(a)(1) through (a)(5).
Otherwise, the commercial nuclear
plant is an interaction-dependentmitigation facility.
(b) Purpose and applicability. The
regulations in §§ 53.725 through 53.830
address areas related to staffing,
training, personnel qualifications, and
human factors engineering for
applicants for or holders of OLs or COLs
under this part. These regulations are
organized as follows:
(1) Sections 53.725 through 53.745
address general requirements for
staffing, training, personnel
qualifications, and human factors
engineering. The regulations within
these sections are applicable to all
applicants for or holders of OLs or COLs
under this part, except where
specifically stated otherwise.
(2) Sections 53.760 through 53.795
address operator and senior operator
licensing requirements. The regulations
within these sections are applicable to
those applicants for or holders of OLs or
COLs under this part for interactiondependent-mitigation facilities that have
not yet certified the permanent
cessation of operations and permanent
removal of fuel from the reactor vessel
as described under § 53.1070.
(3) Sections 53.800 through 53.820
address generally licensed reactor
operator requirements. The regulations
within these sections are in lieu of
§§ 53.760 through 53.795 for those
applicants for or holders of OLs or COLs
under this part for self-reliant-mitigation
facilities that have not yet certified the
permanent cessation of operations and
permanent removal of fuel from the
reactor vessel as described under
§ 53.1070.
(4) Section 53.830 provides general
personnel training requirements. The
regulations within this section are
applicable to all applicants for or
holders of OLs or COLs under this part.
(c) Definitions. When used in
§§ 53.725 through 53.830:
Applicant refers to an applicant for an
operator or senior operator license;
licensee refers to the holder of an
operator, senior operator, or generally
licensed reactor operator license; and
facility licensee refers to the licensee for
the commercial nuclear plant where the

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applicant would be licensed or the
licensee is licensed.
Automation means a device or system
that accomplishes (partially or fully) a
function or task.
Auxiliary operator means any
individual who operates components of
a commercial nuclear plant but does not
manipulate controls or direct the
manipulation of controls of the plant
and is not required to be licensed under
the provisions of this part.
Controls when used with respect to a
nuclear reactor means apparatus and
mechanisms, the manipulation of which
directly affects the reactivity or power
level of the reactor.
Generally licensed reactor operator
means any individual licensed under
the provisions of § 53.810 to manipulate
controls of a self-reliant-mitigation
facility and to direct the licensed
activities of generally licensed reactor
operators.
Interaction-dependent-mitigation
facility means a commercial nuclear
plant design other than one that
demonstrates compliance with the
operating and technical characteristics
defined under § 53.800.
Load following means a commercial
nuclear plant automatically changing its
output to match expected demand in
response to externally originated
instructions or signals.
Operator means any individual
licensed under the provisions of
§§ 53.760 through 53.795 to manipulate
controls of an interaction-dependentmitigation facility.
Performance testing means testing
conducted to verify a simulation
facility’s performance as compared to
actual or predicted reference plant
performance.
Reference plant means the specific
commercial nuclear plant on which a
simulation facility’s configuration,
system control arrangement, and design
data are based. The reference plant may
or may not be constructed.
Self-reliant-mitigation facility means a
commercial nuclear plant design that
demonstrates compliance with the
operating and technical characteristics
defined under § 53.800.
Senior operator means any individual
licensed under the provisions of
§§ 53.760 through 53.795 to manipulate
controls of an interaction-dependentmitigation facility and to direct the
licensed activities of operators.
Simulation facility means an interface
designed to provide a realistic imitation
of the operation of a commercial nuclear
plant used for the administration of
examinations, for training, and/or to
demonstrate compliance with
experience requirements for applicants

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or licensees. A simulation facility may
rely, in whole or part, upon the physical
utilization of the reference plant itself.
Systems approach to training means a
training program that includes the
following five elements:
(1) Systematic analysis of the jobs to
be performed.
(2) Learning objectives derived from
the analysis which describe desired
performance after training.
(3) Training design and
implementation based on the learning
objectives.
(4) Evaluation of trainee mastery of
the objectives during training.
(5) Evaluation and revision of the
training based on the performance of
trained personnel in the job setting.
§ 53.726

Communications.

(a) An applicant or licensee or facility
licensee must submit any
communication or report required by
the regulations contained within
§§ 53.725 through 53.830 and must
submit any application filed under these
regulations to the Commission.
(b) Each licensee that is required to
comply with the requirements of
§§ 53.760 through 53.795 (i.e.,
interaction-dependent-mitigation
facilities) must notify the appropriate
NRC contact within 30 days of the
following in regard to a licensed
operator or senior operator:
(1) Permanent reassignment from the
position for which the licensee has
certified the need for a licensed operator
or senior operator under § 53.775(a)(1);
(2) Termination of any operator or
senior operator; or
(3) Permanent disability or illness as
required under § 55.770 of this chapter.
§ 53.728 Completeness and accuracy of
information.

Information provided to the
Commission by an applicant for an
operator or senior operator license or by
a licensee or information required by
statute or by the Commission’s
regulations, orders, or license
conditions to be maintained by the
applicant or the licensee must be
complete and accurate in all material
respects.
§ 53.730 Defining, fulfilling, and
maintaining the role of personnel in
ensuring safe operations.

Each applicant for or holder of an OL
or COL for a commercial nuclear plant
under this part must comply with the
following:
(a) Human factors engineering design
requirements. The plant design must
reflect state-of-the-art human factors
engineering principles for safe and
reliable performance in all locations that

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human activities are expected for
performing or supporting the continued
availability of plant safety or emergency
response functions.
(b) Human system interface design
requirements. The plant design must
provide for the following to support
operating personnel in monitoring plant
conditions and responding to plant
events:
(1) Features for displaying to
operating personnel a minimum set of
parameters that define the safety status
of the plant and are capable of
displaying both the full range of
important plant parameters and data
trends on demand, as well as indicating
when process limits are being
approached or exceeded;
(2) Automatic indication of the
bypassed and operable status of safety
systems;
(3) Direct indication of SSC status that
relates to the ability of the SSC to
perform its safety function, such as
relief and safety valve position (i.e.,
open or closed) for barriers important to
fulfilling safety functions of with such
devices, and ultimate heat sink and
cooling system status and availability;
(4) Instrumentation to measure,
record, and display key plant
parameters related to the performance of
SSCs and the integrity of barriers
important to fulfilling safety functions
to support operators in monitoring plant
conditions and responding to plant
events. Examples include temperatures
and pressures within important systems
or structures, core or fuel system
conditions (including possible damage
states), temperatures and levels
associated with cooling functions,
combustible gas concentrations,
radiation levels in systems and within
structures, and radioactive effluent
releases;
(5) Leakage control and detection in
the design of systems that pass through
barriers important to fulfilling safety
functions for the release of
radionuclides. An example is an SSC
that penetrates a containment structure
that might contain radioactive materials
that could contribute to the source term
during an accident;
(6) Monitoring of in-plant radiation
and airborne radioactivity as
appropriate for a broad range of normal
operating and accident conditions; and
(7) For self-reliant-mitigation
facilities, the plant design must also
provide the generally licensed reactor
operators with the capability to do the
following:
(i) Receive plant operating data,
including reactor parameters and
information needed for the evaluation of
emergency conditions.

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(ii) Immediately initiate a reactor
shutdown from their location.
(iii) Promptly dispatch operations and
maintenance personnel.
(iv) Immediately implement
responsibilities under the facility
emergency plan, as applicable.
(c) Concept of operations. A concept
of operations that is of sufficient scope
and detail to address the following must
be provided:
(1) Plant goals;
(2) The roles and responsibilities of
operating personnel and automation (or
any combination thereof) that are
responsible for completing plant
functions;
(3) Staffing, qualifications, and
training;
(4) The management of normal
operations;
(5) The management of off-normal
conditions and emergencies;
(6) The management of maintenance
and modifications; and
(7) The management of tests,
inspections, and surveillances.
(d) Functional requirements analysis
and function allocation. A functional
requirements analysis and a function
allocation must be provided that are
sufficient to demonstrate compliance
with the following:
(1) The functional requirements
analysis must address how safety
functions and functional safety criteria
are satisfied, and
(2) The function allocation must
describe how the safety functions will
be assigned to human action,
automation, active safety features,
passive safety features, and/or inherent
safety characteristics.
(e) Operating experience. A program,
during construction and during
operation, as applicable, for evaluating
and applying operating experience must
be developed, implemented, and
maintained.
(f) Staffing plan. A staffing plan must
be developed and comply with the
following:
(1) The staffing plan must include a
description of how engineering
expertise will be available to the onshift operating personnel during all
plant conditions, to assist if they
encounter a situation not covered by
procedures or training. Engineering
expertise includes familiarity with the
operation of the plant for which the
expertise is provided and one of the
following:
(i) A bachelor’s degree in engineering,
engineering technology, or physical
science from an institution accredited
by a U.S. government recognized
accrediting body or equivalent; or
(ii) A Professional Engineer’s license
from a U.S. State or territory.

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(2) Applicants for or holders of OLs or
COLs for interaction-dependentmitigation facilities must include within
their staffing plans a description of how
the proposed numbers, positions, and
qualifications of operators and senior
operators across all modes of plant
operations will be sufficient to ensure
that plant safety functions will be
maintained. This description must be
supported by human factors engineering
analyses and assessments.
(3) Applicants for or holders of OLs or
COLs for self-reliant-mitigation facilities
must include within their staffing plans
a description of how generally licensed
reactor operator staffing that is both
sufficient to continually monitor the
operations of fueled reactors and to
provide for a continuity of
responsibility for facility operations at
all times during the operating phase will
be maintained.
(4) Applicants for or holders of OLs or
COLs under this part must include
within their staffing plans a description
of how the numbers, positions, and
responsibilities of personnel contained
within those plans will adequately
support all necessary functions within
areas such as plant operations,
equipment surveillance and
maintenance, radiological protection,
chemistry control, fire brigades,
engineering, security, and emergency
response.
(5) The staffing plan must be
approved by the NRC as part of its
approval of the OL or COL for the plant.
The approved staffing plan is subject to
the requirements of § 53.1565.
(g) Training, examination, and
proficiency programs. Develop,
implement, and maintain programs that
comply with the following
requirements. These programs must be
approved by the NRC as part of its
approval of the OL or COL for the plant:
(1) For those applicants for or holders
of OLs or COLs for interactiondependent-mitigation facilities:
(i) The operator licensing initial
training program required under
§ 53.780(a);
(ii) The operator licensing initial
examination program required under
§ 53.780(b);
(iii) The operator licensing
requalification program required under
§ 53.780(c); and
(iv) The operator proficiency program
required under § 53.780(g).
(2) For those applicants for or holders
of OLs or COLs for self-reliantmitigation facilities, the generally
licensed reactor operator training,
examination, and proficiency programs
required under § 53.815.

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(3) The operator licensing
requalification programs required under
§ 53.780(c) or § 53.815(b) must be
implemented upon commencing the
administration of initial examinations
under the operator licensing
examination program required under
§ 53.780(b) or § 53.815(b), respectively.
§ 53.735

General exemptions.

The regulations in §§ 53.725 through
53.830 do not require a license for an
individual who—
(a) Under the direction and in the
presence of an operator or senior
operator or a generally licensed reactor
operator, as appropriate, manipulates
the controls of a commercial nuclear
plant as a part of the individual’s
training in a facility licensee’s training
program as approved by the
Commission to qualify for an operator or
senior operator license or a generally
licensed reactor operator license there,
as appropriate, under these regulations;
or
(b) Under the direction and in the
presence of a senior operator or
generally licensed reactor operator, as
appropriate, manipulates the controls of
a commercial nuclear plant to load or
unload the fuel into, out of, or within
the reactor vessel while the reactor is
not operating.

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§ 53.740 Facility licensee requirements—
General.

(a) Facility licensees must
demonstrate compliance with the
requirements of either §§ 53.760 through
53.795 for interaction-dependentmitigation facilities or §§ 53.800 through
53.820 for self-reliant-mitigation
facilities.
(b) The facility licensee must
maintain the staffing complement
described under its approved facility
staffing plan until such time as the
permanent cessation of operations and
permanent removal of fuel from the
reactor vessel has been certified as
described under § 53.1070. The
approved staffing plan is subject to the
requirements of § 53.1565.
(c) Except as provided under § 53.735,
the facility licensee may not permit the
manipulation of the controls of a
commercial nuclear plant by anyone
who is not an operator or senior
operator or generally licensed reactor
operator, as appropriate.
(d) Facility licensees for interactiondependent-mitigation facilities that have
not yet certified the permanent
cessation of operations and permanent
removal of fuel from the reactor vessel
as described under § 53.1070 must
designate senior operators to be

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responsible for supervising the licensed
activities of operators.
(e) Apparatus and mechanisms other
than controls, the operation of which
may affect the reactivity or power level
of a reactor, must be manipulated only
while plant conditions are being
monitored by an individual who is an
operator or senior operator or a
generally licensed reactor operator, as
appropriate.
(f)(1) Load following is permitted if at
least one of the following is
immediately capable of refusing
demands when they could challenge the
safe operation of the plant or when
precluded by the plant equipment
conditions:
(i) The actuation of an automatic
protection system that utilizes setpoints
more conservative than those otherwise
credited for the purposes of reactor
protection; or
(ii) An automated control system; or
(iii) An operator or senior operator or
a generally licensed reactor operator, as
appropriate.
(2) The provisions of paragraph (e) of
this section do not apply during load
following operations.
(g)(1) Facility licensees for
interaction-dependent-mitigation
facilities must have present during
alteration of the core (including fuel
loading or transfer) an individual
holding a senior operator license, or a
senior operator license limited to fuel
handling to directly supervise the
activity and, during this time, the
facility licensee must not assign other
duties to this person.
(2) Facility licensees for self-reliantmitigation facilities must have present
during alteration of the core (including
fuel loading or transfer) an individual
holding a generally licensed reactor
operator license to directly supervise
the activity and, during this time, the
facility licensee must not assign other
duties to this person.
(3) The provisions of paragraphs (g)(1)
and (2) of this section do not apply to
core alterations performed as part of
refueling operations while a facility that
is capable of online refueling is
operating at power.
(h) Facility licensees may take
reasonable action that departs from a
license condition or a technical
specification (contained in a license
issued under this part) in an emergency
when this action is immediately needed
to protect the public health and safety
and no action consistent with license
conditions and technical specifications
that can provide adequate or equivalent
protection is immediately apparent.
Such facility licensee action must be
approved, as a minimum, by a senior

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operator or a generally licensed reactor
operator, as applicable, or, after
certifying the permanent cessation of
operations and permanent removal of
fuel from the reactor vessel as described
under § 53.1070 by a certified fuel
handler, senior operator, or generally
licensed reactor operator, as applicable,
prior to taking the action.
§ 53.745

Operator license requirements.

A person must be authorized by a
license issued by the Commission to
perform the function of an operator,
senior operator, or generally licensed
reactor operator as defined in this part.
§ 53.760

Operator licensing.

(a) Applicability. Sections 53.760
through 53.795 address operator and
senior operator licensing requirements.
The regulations within these sections
are applicable to those applicants for or
holders of OLs or COLs under this part
for interaction-dependent-mitigation
facilities that have not yet certified the
permanent cessation of operations and
permanent removal of fuel from the
reactor vessel as described under
§ 53.1070.
(b) Reserved.
§ 53.765

Medical requirements.

(a) An applicant for an operator or
senior operator license must have a
medical examination by a physician. An
operator or senior operator must have a
medical examination by a physician
every 2 years.
(b) To certify the medical fitness of an
applicant for an operator or senior
operator license, an authorized
representative of the facility licensee
must complete and sign NRC Form 396,
‘‘Certification of Medical Examination
by Facility Licensee,’’ which can be
obtained by writing the Office of the
Chief Information Officer, U.S. Nuclear
Regulatory Commission, Washington,
DC 20555–0001, by calling 301–415–
7232, or by visiting the NRC’s website
at https://www.nrc.gov and selecting
forms from the index found on the home
page, or by other means provided by the
NRC.
(1) Form NRC 396 must certify that a
physician has conducted the medical
examination of the applicant as required
in paragraph (a) of this section.
(2) When the medical certification
requests a conditional license based on
medical evidence, the medical evidence
must be submitted on NRC Form 396 to
the Commission to enable the
Commission to make a determination in
accordance with § 53.775(b).
(c) The facility licensee must
document and maintain the results of
medical qualifications data, test results,

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and each operator’s or senior operator’s
medical history for the current license
period and provide the documentation
to the Commission upon request. The
facility licensee must retain this
documentation while an individual
performs the functions of an operator or
senior operator.
§ 53.770 Incapacitation because of
disability or illness.

If, during the term of the operator or
senior operator license, the licensee
develops a permanent physical or
mental condition that causes the
licensee to fail to demonstrate
compliance with the requirements of
§ 53.775(b)(1)(i), the facility licensee
must notify the Commission within 30
days of learning of the diagnosis. For
conditions for which a conditional
license (as described in § 53.775(b)) is
requested, the facility licensee must
provide medical certification on Form
NRC 396 to the Commission (as
described in § 53.765(b)).

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§ 53.775 Applications for operators and
senior operators.

(a) How to apply. (1) The applicant for
an operator or senior operator license
must—
(i) Complete NRC Form 398,
‘‘Personal Qualification Statement—
Licensee,’’ which can be obtained by
writing the Office of the Chief
Information Officer, U.S. Nuclear
Regulatory Commission, Washington,
DC 20555–0001, by calling 301–415–
5877, or by visiting the NRC’s website
at https://www.nrc.gov and selecting
forms from the index found on the home
page, or by other means provided by the
NRC;
(ii) File an original of NRC Form 398,
or an equivalent electronic submittal,
together with the information required
in paragraphs (a)(1)(iii) and (a)(1)(iv) of
this section, with the appropriate
Regional Administrator.
(iii) Provide evidence that the
applicant, as a trainee, has successfully
demonstrated competence in
manipulating the controls of either the
facility for which a license is sought or
a simulation facility that demonstrates
compliance with the requirements of
§ 53.780(e). For operators applying for a
senior operator license, certification that
the operator has successfully operated
the controls of the facility as an operator
will be accepted; and
(iv) Provide certification by the
facility licensee of medical condition
and general health on Form NRC 396, to
comply with § 53.765.
(2) The Commission may at any time
after the application has been filed, and
before the license has expired, require

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further information under oath or
affirmation to enable it to determine
whether to grant or deny the application
or whether to revoke, modify, or
suspend the license.
(3) An applicant whose application
has been denied because of a medical
condition or their general health may
submit a further medical report at any
time as a supplement to the application.
(4) Each application and statement
must contain complete and accurate
disclosure as to all matters required to
be disclosed. The applicant must sign
statements required by paragraphs
(a)(1)(i) and (a)(1)(ii) of this section.
(b) Disposition of an initial
application. (1) License approval. The
Commission will approve an initial
application if it finds that the following
criteria are met:
(i) Health. The applicant’s medical
condition and general health will not
adversely affect the performance of
assigned operator or senior operator job
duties or cause operational errors
endangering public health and safety.
The Commission will base its finding
upon the certification by the facility
licensee as detailed in § 53.765(b).
(ii) Examination. The applicant has
passed the requisite examination in
accordance with § 53.780(b). The
examination determines whether the
applicant for an operator’s or senior
operator’s license has learned to operate
a facility competently and safely, and
additionally, in the case of a senior
operator, whether the applicant has
learned to supervise the licensed
activities of operators competently and
safely.
(2) Conditional license. If an
applicant’s general medical condition
does not demonstrate compliance with
the minimum standards under
§ 53.775(b)(1)(i) of this section, the
Commission may approve the
application and include conditions in
the license to accommodate the medical
condition. The Commission will
consider the recommendations and
supporting evidence of the facility
licensee and of the examining physician
(provided on Form NRC 396) in arriving
at its decision.
(c) Re-applications. (1) An applicant
whose application for a license has been
denied because of failure to pass the
examination may file a new application.
The application must be submitted on
Form NRC 398 and include a statement
signed by an authorized representative
of the facility licensee by whom the
applicant will be employed that states
in detail the extent of the applicant’s
additional training and remediation
since the denial and certifies that the
applicant is ready for re-examination.

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(2) An applicant who has passed a
portion of the examination and failed
another may request in a new
application on Form NRC 398 to be
excused from re-examination on the
portions of the examination that the
applicant has passed. The Commission
may in its discretion grant the request
if it determines that sufficient
justification is presented.
§ 53.780 Training, examination, and
proficiency program.

(a) Operator licensing initial training
program. (1) A program that is based
upon a systems approach to training, as
defined by § 53.725(b), must be utilized
for the training of applicants for
operator and senior operator licenses.
The program must ensure that
applicants at the facility will possess the
knowledge, skills, and abilities
necessary to protect the public health
and maintain those plant safety
functions specific to the facility design.
The program must be approved by the
Commission prior to its use for training
applicants, as described under
§ 53.730(g). The approved operator
licensing initial training program is
subject to the requirements of § 53.1565.
(2) The facility licensee must
maintain operator licensing initial
training program records documenting
the initial operator licensing training
administered and completed by each
applicant. The facility licensee must
retain these records during the period in
which any trainees subsequently remain
licensed as operators or senior operators
at the facility.
(b) Operator licensing initial
examination program. (1) The facility
licensee must establish and implement
an examination program for testing a
representative sample of the knowledge,
skills, and abilities needed to safely
perform operator and senior operator
duties, to include both the examination
methods and criteria to be used to assess
passing performance. The program must
provide for valid and reliable
examinations and be approved by the
Commission prior to its use for
examining applicants, as described
under § 53.730(g). The approved
operator licensing initial examination
program is subject to the requirements
of § 53.1565.
(2) The facility licensee must submit
prepared examinations to the
Commission for review and approval in
advance of their administration.
(3) The Commission will either
administer an approved examination or
allow the facility licensee to administer
the examination. The facility licensee
must ensure that sufficient advance
notification is provided to the

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Commission to either administer the
examination or allow for a
representative of the Commission to be
afforded the opportunity to be present
when the facility licensee administers
the examination.
(4) Graded examination
documentation for each applicant must
be promptly provided to the
Commission for review in making
operator licensing decisions.
(5) The facility licensee must
maintain operator licensing initial
examination program records
documenting the participation of each
operator and senior operator applicant
in the initial examination. The records
must contain copies of examinations
administered, the answers given by the
applicant, documentation of the grading
of examinations, and documentation of
any additional training administered in
areas in which an applicant exhibited
deficiencies. The facility licensee must
retain these records during the period in
which the associated operators or senior
operators remain licensed at the facility.
(c) Operator licensing requalification
program. (1) A program based upon a
systems approach to training, as defined
by § 53.725(b), must be utilized for the
continuing training of operators and
senior operators.
(i) The program must ensure that
operators and senior operators at the
facility maintain the knowledge, skills,
and abilities necessary to protect the
public health and maintain those plant
safety functions specific to the facility
design. The program must be conducted
for a continuous period not to exceed 24
months in duration.
(ii) The program must be approved by
the Commission prior to its use for
continuing training, as described under
§ 53.730(g). The approved operator
licensing requalification program is
subject to the requirements of § 53.1565.
(2) The following requirements apply
to operator licensing requalification
programs:
(i) The facility licensee must propose
a requalification examination program
for testing, for each requalification
period, a sample of the topics included
under the systems approach to training,
to include both the examination
methods and criteria to be used to assess
passing performance. The program must
provide for valid and reliable
examinations and be approved by the
Commission prior to its use for
examining operators and senior
operators, as described under
§ 53.730(g). The approved
requalification examination program is
subject to the requirements of § 53.1565.

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(ii) The following requirements apply
to the requalification examination
program:
(A) The facility licensee must make
prepared requalification examinations
available to the Commission for review.
(B) The facility licensee must ensure
that a representative of the Commission
is afforded the opportunity to be present
during requalification examination
administration.
(C) The facility licensee must ensure
that each operator and senior operator is
administered a complete requalification
examination on a periodicity not to
exceed 24 months. Additionally, the
facility licensee must ensure that any
licensed operator or senior licensed
operator who either demonstrates
unsatisfactory performance on, or fails
to complete, the biennial requalification
examination is removed from the
performance of licensed operator and
senior licensed operator duties until
such time that any necessary remedial
training has been completed and a
retake examination has been passed.
(D) The facility licensee must
promptly provide a summary of
examination results for each operator
and senior operator following the
completion of the requalification
examination.
(3) The facility licensee must
maintain operator licensing
requalification program records
documenting the participation of each
operator and senior operator in the
requalification program. The records
must contain copies of examinations
administered, the answers given by the
operator or senior operator,
documentation of the grading of
examinations, and documentation of
any additional training administered in
areas in which an operator or senior
operator exhibited deficiencies. The
facility licensee must retain these
records until the operator’s or senior
operator’s license is renewed.
(d) Examination integrity. Applicants,
operators and senior operators, and
facility licensees must not engage in any
activity that compromises the integrity
of any application or examination
required by §§ 53.760 through 53.795.
The integrity of an examination is
considered compromised if any activity,
regardless of intent, affected, or, but for
detection, could have affected the
equitable and consistent administration
of the examination. This includes
activities related to the preparation and
certification of applications and all
activities related to the preparation,
administration, and grading of
examinations required by §§ 53.760
through 53.795.

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87065

(e) Simulation facilities. (1) This
section addresses the use of a
simulation facility for the
administration of examinations, for
training, or to demonstrate compliance
with experience requirements for
applicants for operator and senior
operator licenses.
(2) Simulation facilities used for
training purposes, for demonstrating
compliance with experience
requirements, or for the conduct of
examinations under § 53.780(b) and (c)
must demonstrate compliance with the
following criteria as they relate to the
facility licensee’s reference plant:
(i) The simulation facility must be of
sufficient scope and fidelity for
individuals to acquire and demonstrate
the necessary knowledge, skills, and
abilities to safely perform operator and
senior operator duties.
(ii) The simulation facility must
utilize models relating to nuclear,
thermal-hydraulic, and other applicable
design-specific characteristics that
either replicate the most recent fuel load
in the reference commercial nuclear
plant or, prior to initial fuel load (or, for
a fueled manufactured reactor, prior to
initiating the physical removal of any
one of the independent physical
mechanisms to prevent criticality
required under § 53.620(d)(1)), replicate
the intended initial fuel load for the
reference commercial nuclear plant,
with the exception of those portions of
the simulation facility that utilize the
reference plant itself.
(iii) Simulation facility fidelity must
be demonstrated so that significant
control manipulations are completed
without procedural exceptions,
simulator performance exceptions, or
deviation from the approved training
scenario sequence.
(3) Facility licensees that maintain a
simulation facility that has been
approved by the Commission for
training purposes, demonstrating
compliance with experience
requirements, or the conduct of
examinations under § 53.780(b) and (c)
for the facility licensee’s reference plant
must:
(i) Conduct performance testing
throughout the life of the simulation
facility in a manner sufficient to ensure
that paragraph (e)(2) of this section is
met;
(ii) Retain the results of performance
testing for 4 years after the completion
of each performance test or until
superseded by updated test results;
(iii) Promptly correct modeling and
hardware discrepancies and
discrepancies identified from scenario
validation and from performance testing
or provide justification as to why the

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presence of such discrepancies will not
adversely affect simulator performance
with respect to the criteria of paragraph
(e)(2) of this section;
(iv) Make the results of any
uncorrected performance test failures
that may exist at the time of the initial
license examination or requalification
examination available for NRC review,
prior to or concurrent with preparations
for each initial license examination or
requalification examination; and
(v) Maintain the provisions for license
application and examination integrity
consistent with § 53.780(d).
(4) A simulation facility must
demonstrate compliance with the
requirements of paragraphs (e)(2) and
(e)(3) of this section for the Commission
to accept the simulation facility for
conducting initial examinations as
described in § 53.780(b), requalification
training as described in § 53.780(c), or
performing control manipulations that
affect reactivity to establish eligibility
for an operator or senior operator
license as described in § 53.775(a).
(f) Waiver of examination
requirement. On application, the
Commission may waive any or all of the
requirements for an examination if it
finds that the applicant has
demonstrated the required knowledge,
skills, and abilities to safely operate the
plant, and is capable of continuing to do
so. The Commission may make such a
finding based on demonstration of the
following:
(1) Operating experience at a
comparable facility;
(2) Proof of the applicant’s past
competent and safe performance; and
(3) Proof of the applicant’s current
qualifications.
(g) Proficiency. The facility licensee
must develop, implement, and maintain
a proficiency program to ensure that
operators and senior operators will
actively perform the functions of an
operator or senior operator, respectively,
as needed to maintain proficiency with
on-shift duties and familiarity with
plant status. This program must include
those steps that will be taken to reestablish proficiency when it cannot be
maintained. This program must be
approved by the Commission as part of
its approval of the OL or COL for the
plant. The approved proficiency
program is subject to the requirements
of § 53.1565.
(h) Records. Each record required by
this section must be legible throughout
the retention period specified by each
Commission regulation. The record may
be the original, a reproduced copy, or an
electronic copy provided that the copy
is authenticated by authorized
personnel.

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§ 53.785 Conditions of operator and senior
operator licenses.

Each operator and senior operator
license contains and is subject to the
following conditions whether stated in
the license or not:
(a) Neither the license nor any right
under the license may be assigned or
otherwise transferred.
(b) The license is limited to the
facility for which it is issued.
(c) The license is limited to those
controls of the facility or facilities
specified in the license.
(d) The license is subject to, and the
licensee must observe, all applicable
rules, regulations, and orders of the
Commission.
(e) The licensee must maintain or reestablish proficiency in accordance with
the facility licensee’s Commissionapproved proficiency program required
under § 53.780(g).
(f) The licensee must be subject to the
facility’s Commission-approved
operator licensing requalification and
requalification examination programs
required under § 53.780(c).
(g) The licensee must have a biennial
medical examination as described by
§ 53.765.
(h) The licensee must notify the
Commission within 30 days about a
conviction for a felony.
(i) The licensee must not consume or
ingest alcoholic beverages within the
protected area of commercial nuclear
plants. The licensee must not use,
possess, or sell any illegal drugs. The
licensee must not perform activities
authorized by a license issued under
this part while under the influence of
alcohol or any prescription, over-thecounter, or illegal substance that could
adversely affect his or her ability to
safely and competently perform his or
her licensed duties. For the purpose of
this paragraph, with respect to alcoholic
beverages and drugs, the term ‘‘under
the influence’’ means the licensee
exceeded, as evidenced by a confirmed
test result, the lower of the cutoff levels
for drugs or alcohol contained in 10 CFR
part 26, or as established by the facility
licensee. The term ‘‘under the
influence’’ also means the licensee
could be mentally or physically
impaired as a result of substance use
including prescription and over-thecounter drugs, as determined under the
provisions, policies, and procedures
established by the facility licensee for
its fitness-for-duty program, in such a
manner as to adversely affect his or her
ability to safely and competently
perform licensed duties.
(j) Each licensee must participate in
the drug and alcohol testing programs as
required under 10 CFR part 26.

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(k) The licensee must comply with
any other conditions that the
Commission may impose to protect
health or to minimize danger to life or
property.
§ 53.790 Issuance, modification, and
revocation of operator and senior operator
licenses.

(a) Issuance of operator and senior
operator licenses. If the Commission
determines that an applicant for an
operator license or a senior operator
license demonstrates compliance with
the requirements of the Atomic Energy
Act of 1954, as amended, (the Act) and
its regulations, it will issue a license in
the form and containing any conditions
and limitations it considers appropriate
and necessary.
(b) Modification and revocation of
operator and senior operator licenses.
(1) The terms and conditions of all
operator and senior operator licenses are
subject to amendment, revision, or
modification by reason of rules,
regulations, or orders issued in
accordance with the Act or any
amendments thereto.
(2) Any license may be revoked,
suspended, or modified, in whole or in
part—
(i) For any material false statement in
the application or in any statement of
fact required under section 182 of the
Act;
(ii) Because of conditions revealed by
the application or statement of fact or
any report, record, inspection, or other
means that would warrant the
Commission to refuse to grant a license
on an original application;
(iii) For willful violation of, or failure
to observe, any of the terms and
conditions of the Act or the license, or
of any rule, regulation, or order of the
Commission;
(iv) For any conduct determined by
the Commission to be a hazard to safe
operation of the facility; or
(v) For the sale, use, or possession of
illegal drugs, or refusal to participate in
the facility drug and alcohol testing
program, or a confirmed positive test for
drugs, drug metabolites, or alcohol in
violation of the conditions and cutoff
levels established by § 53.785(i) or the
consumption of alcoholic beverages
within the protected area of commercial
nuclear plants, or a determination of
unfitness for scheduled work as a result
of the consumption of alcoholic
beverages.
§ 53.795 Expiration and renewal of
operator and senior operator licenses.

(a) Expiration. (1) Each operator
license and senior operator license
expires 6 years after the date of

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issuance, upon termination of
employment with the facility licensee,
or upon determination by the facility
licensee that the licensed individual no
longer needs to maintain a license.
(2) If a licensee files an application for
renewal or an upgrade of an existing
license on Form NRC 398 at least 30
days before the expiration of the
existing license, it does not expire until
disposition of the application for
renewal or for an upgraded license has
been finally determined by the
Commission. Filing by mail will be
deemed to be complete at the time the
application is postmarked
(b) Renewal. (1) The applicant for
renewal of an operator license or senior
operator license must—
(i) Complete and sign Form NRC 398
and include the number of the license
for which renewal is sought.
(ii) File an original of NRC Form 398
as specified in § 53.775.
(iii) Provide written evidence of the
applicant’s experience under the
existing license and the approximate
number of hours that the licensee has
operated the facility.
(iv) Provide a statement by an
authorized representative of the facility
licensee that during the effective term of
the current license the applicant has
satisfactorily completed the
requalification program for the facility
for which operator or senior operator
license renewal is sought.
(v) Provide evidence that the
applicant has discharged the license
responsibilities competently and safely.
The Commission may accept as
evidence of the applicant’s having met
this requirement a certificate of an
authorized representative of the facility
licensee or holder of an authorization by
which the licensee has been employed.
(vi) Provide certification by the
facility licensee of medical condition
and general health on Form NRC 396, to
comply with § 53.765.
(2) The license will be renewed if the
Commission finds that—
(i) The medical condition and the
general health of the licensee continue
to be such as not to cause operational
errors that endanger public health and
safety. The Commission will base this
finding upon the certification by the
facility licensee as described in
§ 53.765(b).
(ii) The licensee—
(A) Is capable of continuing to
competently and safely assume licensed
duties;
(B) Has successfully completed a
requalification program that has been
approved by the Commission as
required by § 53.780(c); and

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(C) Has passed the requalification
examinations as required by § 53.780(c).
(iii) There is a continued need for an
operator to operate or for a senior
operator to supervise operators at the
facility designated in the application.
(iv) The past performance of the
licensee has been satisfactory to the
Commission. In making its finding, the
Commission will include in its
evaluation information such as notices
of violations or letters of reprimand in
the licensee’s docket.
§ 53.800 Facility licensees for self-reliantmitigation facilities.

(a) A commercial nuclear plant is a
self-reliant-mitigation facility if the NRC
determined as part of its approval of the
OL or COL for that plant that its design
demonstrates compliance with criteria
(a)(1) though (a)(5) of this section. A
self-reliant-mitigation facility is of a
class, based upon the similarity of
operating and technical characteristics
of the plants in the class, such that its
licensee must comply with the
requirements of §§ 53.800 through
53.820 in lieu of those in §§ 53.760
through 53.795.
(1) The safety performance criteria of
§§ 53.210 and 53.220 and, if applicable,
any alternative criteria used in
accordance with § 53.470, must be met
without reliance upon human action for
credited event mitigation.
(2) The results of a probabilistic risk
analysis must demonstrate that the
evaluation criteria for the events
analyzed in accordance with § 53.450
will be met without reliance on human
actions to achieve acceptable event
mitigation.
(3) The functional requirements
analysis and function allocation
performed under § 53.730(d) must
demonstrate that functions required for
safety are not reliant upon credited
human action.
(4) The plant response to events
analyzed under § 53.450 must rely
exclusively on safety features and
characteristics that will neither be
rendered unavailable by credible human
errors of commission or omission nor
credibly require manual human
operation in response to equipment
failures. Compliance with this
paragraph may be achieved through the
use of SSCs that function through
inherent characteristics or that have
engineered protections against human
failures.
(5) The plant design must provide for
a layered defense-in-depth approach
that is not dependent upon any single
barrier or credited human action.
(b) [Reserved]

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§ 53.805 Facility licensee requirements
related to generally licensed reactor
operators.

(a) Licensees for self-reliantmitigation facilities that have not yet
certified the permanent cessation of
operations and permanent removal of
fuel from the reactor vessel as described
under § 53.1070 must demonstrate
compliance with the following
requirements:
(1) Ensure that, in addition to being
qualified to perform those items
identified by the facility-specific
systems approach to training conducted
under § 53.815, generally licensed
reactor operators are qualified to safely
and competently—
(i) Perform administrative tasks,
including compliance with technical
specifications, and perform operability
determinations;
(ii) Implement maintenance and
configuration controls;
(iii) Comply with radioactive release
limitations;
(iv) Understand plant operating data,
including reactor parameters, and
evaluate emergency conditions;
(v) Initiate a reactor shutdown from
necessary locations;
(vi) Dispatch and direct operations
and maintenance personnel;
(vii) Implement any applicable
responsibilities under the facility
emergency plan; and
(viii) Make required notifications to
local, State, participating Tribal and
Federal authorities.
(2) Develop, implement, and maintain
facility technical specifications that
provide the necessary administrative
controls to ensure the implementation
of these requirements.
(3) Develop, implement, and maintain
the generally licensed reactor operator
training, examination, and proficiency
programs required under § 53.815.
(4) Ensure that generally licensed
reactor operators are subject to the
facility’s generally licensed reactor
operator training, examination, and
proficiency programs required under
§ 53.815. Ensure that generally licensed
reactor operators are subject to and
comply with the applicable
programmatic requirements for plant
personnel required under 10 CFR parts
26 and 73. An individual that is not in
compliance with any of these programs
is not qualified to be in a position that
may involve the manipulation of the
controls of the commercial nuclear
plant.
(5) Report annually to the NRC the
identity of all generally licensed reactor
operators at the commercial nuclear
plant, including all additions and
deletions since the previous report.

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(6) Ensure that the facility design
continues to meet the criteria of
§ 53.800.
(b) [Reserved]

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§ 53.810 Generally licensed reactor
operators.

(a) A general license to manipulate
the controls of a self-reliant-mitigation
facility and to direct the licensed
activities of generally licensed reactor
operators is hereby issued to any
individual employed in a position that
may involve the manipulation of the
controls of that self-reliant-mitigation
facility and who observes the
restrictions of this section.
(b) A generally licensed reactor
operator must comply with the
operating procedures and other
conditions specified in the license
authorizing operation of the facility.
(c) The general license is limited to
the facility or facilities at which the
operator is employed.
(d) The Commission will suspend the
general license on an individual basis
for violations of any provision of the Act
or any rule or regulation issued
thereunder whenever the Commission
deems such suspension desirable,
including—
(1) For willful violation of, or failure
to observe, any of the terms and
conditions of the Act or the general
license, or of any rule, regulation, or
order of the Commission;
(2) For any conduct determined by the
Commission to be a hazard to safe
operation of the facility; or
(3) For the sale, use, or possession of
illegal drugs, or refusal to participate in
the facility drug and alcohol testing
program, or a confirmed positive test for
drugs, drug metabolites, or alcohol in
violation of the conditions and cutoff
levels established by § 53.810(f) or the
consumption of alcoholic beverages
within the protected area of commercial
nuclear plants, or a determination of
unfitness for scheduled work as a result
of the consumption of alcoholic
beverages.
(e) The Commission may require
information from a generally licensed
reactor operator to determine whether a
general license should be revoked or
suspended with respect to that operator.
(f) The generally licensed reactor
operator must not consume or ingest
alcoholic beverages within the protected
area of commercial nuclear plants. The
generally licensed reactor operator must
not use, possess, or sell any illegal
drugs. The generally licensed reactor
operator must not perform activities
requiring a general license while under
the influence of alcohol or any
prescription, over-the-counter, or illegal

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substance that could adversely affect his
or her ability to safely and competently
perform these activities. For the purpose
of this paragraph, with respect to
alcoholic beverages and drugs, the term
‘‘under the influence’’ means the
generally licensed reactor operator
exceeded, as evidenced by a confirmed
test result, the lower of the cutoff levels
for drugs or alcohol contained in 10 CFR
part 26, or as established by the facility
licensee. The term ‘‘under the
influence’’ also means the generally
licensed reactor operator could be
mentally or physically impaired as a
result of substance use including
prescription and over-the-counter drugs,
as determined under the provisions,
policies, and procedures established by
the facility licensee for its fitness-forduty program, in such a manner as to
adversely affect his or her ability to
safely and competently perform
generally licensed reactor operator
duties.
(g) The generally licensed reactor
operator must notify the Commission
within 30 days about a conviction for a
felony.
§ 53.815 Generally licensed reactor
operator training, examination, and
proficiency programs.

(a) Applicability. The requirements of
this section apply to each licensee of a
self-reliant-mitigation facility that has
not yet certified the permanent
cessation of operations and permanent
removal of fuel from the reactor vessel
as described under § 53.1070.
(b) Requirements. (1) The licensee
must develop, implement, and maintain
training and examination programs that
demonstrate compliance with the
requirements of paragraphs (b)(2) and
(3) of this section.
(2) The training program must provide
for both the initial and continuing
training of generally licensed reactor
operators and be derived from a systems
approach to training as defined in this
part.
(3)(i) The training program must
incorporate the instructional
requirements necessary to provide
qualified generally licensed reactor
operators to operate and maintain the
facility in a safe manner in all modes of
operation. The training program must
comply with the facility license,
including all technical specifications
and applicable regulations. The facility
licensee must periodically evaluate and
revise the training program as
appropriate to reflect industry
experience and relevant changes,
including changes to the facility,
procedures, regulations, and quality
assurance (QA) requirements. Facility

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licensee management must periodically
review the training program for
effectiveness.
(ii) The training program must ensure
that generally licensed reactor operators
have and maintain the necessary
knowledge, skills, and abilities.
(iii) The training program must
include the generally licensed reactor
operator manipulating the controls of
either the facility or a simulation facility
that demonstrates compliance with the
requirements of § 53.815(e).
(iv) The training program must
include an initial examination program
for testing a representative sample of the
knowledge, skills, and abilities needed
to safely perform generally licensed
reactor operator duties, to include both
the examination methods and criteria to
be used to assess passing performance.
The facility licensee must provide the
opportunity for a representative of the
Commission to be present during initial
examination administration.
(v) The training program must include
a requalification examination program
for testing a sample of the topics
included under the systems approach to
training, to include the examination
methods and criteria to be used to assess
passing performance. The
requalification examination program
must specify an appropriate periodicity
for administering a complete
requalification examination to each
generally licensed reactor operator, and
the facility licensee must provide the
opportunity for a representative of the
Commission to be present during
requalification examination
administration.
(A) The facility licensee must ensure
that any generally licensed reactor
operator who either demonstrates
unsatisfactory performance on, or fails
to complete, the requalification
examination is removed from the
performance of generally licensed
reactor operator duties until such time
that any necessary remedial training has
been completed and a retake
examination has been passed.
(B) [Reserved]
(vi) The training program must be
approved by the Commission prior to its
use, as described under § 53.730(g). The
examination program must provide for
valid and reliable examinations and
must be approved by the Commission
prior to their use, as described under
§ 53.730(g). The approved programs are
subject to the requirements of § 53.1565.
(c) Records. The following is required
regarding the documentation of the
generally licensed reactor operator
training and examination programs:
(1) Sufficient records must be
maintained by the facility licensee to

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maintain the integrity of the programs
and kept available for NRC inspection to
verify the adequacy of the programs.
(2) The facility licensee must
maintain records documenting the
participation of each generally licensed
reactor operator in the training and
examination programs. The records
must contain copies of examinations
administered, the answers given by the
generally licensed reactor operator,
documentation of the grading of
examinations, and documentation of
any additional training administered in
areas in which a generally licensed
reactor operator exhibited deficiencies.
The facility licensee must retain these
records while the associated generally
licensed reactor operators remain
employed at the facility.
(3) Each record required by this part
must be legible throughout the retention
period. The record may be the original,
a reproduced copy, or an electronic
copy provided that the copy is
authenticated by authorized personnel.
(d) Examination integrity. Generally
licensed reactor operators and facility
licensees must not engage in any
activity that compromises the integrity
of any examination conducted under the
generally licensed reactor operator
training and examination programs. The
integrity of an examination is
considered compromised if any activity,
regardless of intent, affected, or, but for
detection, could have affected the
equitable and consistent administration
of the examination. This includes all
activities related to the preparation,
administration, and grading of
examinations.
(e) Simulation facilities. (1)
Simulation facilities used for training
purposes, for maintaining proficiency,
or for the conduct of examinations must
demonstrate compliance with the
following criteria as they relate to the
facility licensee’s reference plant:
(i) The simulation facility must be of
sufficient scope and fidelity for
individuals to acquire and demonstrate
the necessary knowledge, skills, and
abilities to safely perform generally
licensed reactor operator duties.
(ii) The simulation facility must
utilize models relating to nuclear,
thermal-hydraulic, and other applicable
design-specific characteristics that
either replicate the most recent fuel load
in the reference commercial nuclear
plant or, prior to initial fuel load (or, for
a fueled manufactured reactor, prior to
initiating the physical removal of any
one of the independent physical
mechanisms to prevent criticality
required under § 53.620(d)(1)), replicate
the intended initial fuel load for the
reference commercial nuclear plant,

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with the exception of those portions of
the simulation facility that utilize the
reference plant itself.
(iii) Simulator fidelity must be
demonstrated so that significant control
manipulations are completed without
procedural exceptions, simulator
performance exceptions, or deviation
from the approved training scenario
sequence.
(2) Facility licensees that maintain a
simulation facility for training purposes,
for maintaining proficiency, or for the
conduct of examinations must—
(i) Conduct performance testing
throughout the life of the simulation
facility in a manner sufficient to ensure
that paragraph (e)(1) of this section is
met;
(ii) Retain the results of performance
testing for 4 years after the completion
of each performance test or until
superseded by updated test results;
(iii) Promptly correct modeling and
hardware discrepancies and
discrepancies identified from scenario
validation and from performance testing
or provide justification for why the
presence of such discrepancies will not
adversely affect the criteria of paragraph
(e)(1) of this section;
(iv) Make the results of any
uncorrected performance test failures
that may exist at the time of an
inspection available for NRC review;
and
(v) Maintain the provisions for
examination integrity consistent with
§ 53.815(d).
(f) Waiver of examination
requirement. The facility licensee may
waive any or all of the requirements for
an examination in accordance with the
facility licensee’s Commission-approved
generally licensed reactor operator
training and examination programs.
(g) Proficiency. The facility licensee
must develop, implement, and maintain
a proficiency program to allow generally
licensed reactor operators to maintain
proficiency regarding position functions
and familiarity with plant status. This
program must include those steps that
will be taken in order to re-establish
proficiency when it cannot be
maintained.
§ 53.820 Cessation of individual
applicability.

The general license ceases to be
applicable on an individual basis once
a generally licensed reactor operator is
no longer being employed in a position
that may involve the manipulation of
the controls of the self-reliant mitigation
facility.

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§ 53.830 Training and qualification of
commercial nuclear plant personnel.

(a) This section addresses personnel
training requirements. The regulations
within this section are applicable to all
applicants for or holders of OLs or COLs
under this part.
(b) Prior to initial fuel load (or, for a
fueled manufactured reactor, prior to
initiating the physical removal of any
one of the independent physical
mechanisms to prevent criticality
required under § 53.620(d)(1)), each
holder of an operating or COL under
this part must, with sufficient time to
provide trained and qualified personnel
to operate the facility, establish,
implement, and maintain a training
program that demonstrates compliance
with the requirements of paragraphs (c)
and (d) of this section.
(c) The training program must be
derived from a systems approach to
training as defined in this part and must
provide, at a minimum, for the training
and qualification of the following
categories of commercial nuclear plant
personnel:
(1) Supervisors (e.g., shift
supervisors);
(2) Technicians (e.g., maintenance,
chemistry, and radiological); and
(3) Other appropriate operating
personnel (e.g., auxiliary operators,
certified fuel handlers, and individuals
who provide engineering expertise to
on-shift operating personnel).
(d) The training program must
incorporate the instructional
requirements necessary to provide
qualified personnel to operate
components of a commercial nuclear
plant and maintain the facility in a safe
manner in all modes of operation. The
training program must be developed to
be in compliance with the facility
license, including all technical
specifications and applicable
regulations.
(1) The training program must be
periodically evaluated and revised as
appropriate to reflect industry
experience and relevant changes,
including changes to the facility,
procedures, regulations, and QA
requirements. The training program
must be periodically reviewed by
facility licensee management for
effectiveness.
(2) Sufficient records must be
maintained by the facility licensee to
maintain program integrity and kept
available for NRC inspection to verify
the adequacy of the training program.
§ 53.845

Programs.

(a) The required plant programs under
this part must include but are not
necessarily limited to the programs

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described in the following sections of
this subpart. Licensees may combine,
separate, and otherwise organize
programs and related documents as
appropriate for the technologies and
organizations associated with the
commercial nuclear plant.
(b) In addition to the programs
described in the following sections,
programs must be provided for each
commercial nuclear plant, if necessary,
to ensure that the performance of design
features and human actions are
consistent with the analyses performed
under §§ 53.450 and 53.730 and that the
plant will demonstrate compliance with
the safety criteria defined in §§ 53.210
and 53.220.

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§ 53.850

Radiation protection.

(a) Each holder of an OL or COL
under this part must develop,
implement, and maintain a Radiation
Protection Program for operations that is
commensurate with the scope and
extent of licensed activities under this
part and includes measures for limiting
and monitoring radioactive plant
effluents and limiting and monitoring
the dose to individuals working with
radioactive materials in accordance with
10 CFR part 20.
(b) Each holder of an OL or COL
under this part must develop,
implement, and maintain a program for
the control of radioactive effluents and
for keeping the doses to members of the
public from radioactive effluents as low
as is reasonably achievable and for
environmental monitoring. The program
must be contained in an Offsite Dose
Calculations Manual, must be
implemented by procedures, and must
include remedial actions to be taken
whenever the program limits are
exceeded. The Offsite Dose Calculations
Manual must—
(1) Contain the methodology and
parameters used in the calculation of
offsite doses resulting from radioactive
gaseous and liquid effluents, in the
calculation of gaseous and liquid
effluent monitoring alarm and trip
setpoints, and in the conduct of the
radiological environmental monitoring
program; and
(2) Contain the radioactive effluent
controls and radiological environmental
monitoring activities, and descriptions
of the information that should be
included in the Annual Radiological
Environmental Operating and
Radioactive Effluent Release Reports
required by § 53.1645.
(c) Each holder of an OL or COL
under this part must develop,
implement, and maintain a Process
Control Program that identifies the
administrative and operational controls

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for solid radioactive waste processing,
process parameters, and surveillance
requirements sufficient to ensure
compliance with the requirements of 10
CFR part 20, 10 CFR part 61, and 10
CFR part 71.
§ 53.855

Emergency preparedness.

(a) Each holder of an OL or COL
under this part must have an emergency
response plan that must contain
information needed to demonstrate
compliance with either the
requirements in § 50.160 of this chapter
or the requirements in appendix E to
part 50 and the planning standards of
§ 50.47(b) of this chapter.
(b) No initial OL, initial COL, or early
site permit that includes complete and
integrated emergency plans will be
issued under this part unless a finding
is made by the NRC, in accordance with
§ 50.47 of this chapter, that there is
reasonable assurance that adequate
protective measures can and will be
taken in the event of a radiological
emergency.
§ 53.860

Security programs.

(a) Physical Protection Program. Each
holder of an OL or COL under this part
must develop, implement, and maintain
a physical protection program under the
following requirements:
(1) The licensee must implement
security requirements for the protection
of special nuclear material based on the
type, enrichment, and quantity in
accordance with 10 CFR part 73, as
applicable, and implement security
requirements for the protection of
Category 1 and Category 2 quantities of
radioactive material in accordance with
10 CFR part 37, as applicable; and
(2) The licensee must demonstrate
compliance with the provisions set forth
in either §§ 73.55 or 73.100 of this
chapter, unless the licensee
demonstrates compliance with the
following criterion:
(i) The radiological consequences
from a design-basis threat-initiated
event involving the loss of engineered
systems for decay heat removal and
possible breaches in physical structures
surrounding the reactor, spent fuel, and
other inventories of radioactive
materials result in offsite doses below
the values in § 53.210.
(ii) The applicant must perform a sitespecific analysis, including
identification of target sets, to
demonstrate that the criterion in
§ 53.860(a)(2)(i) is satisfied. The analysis
must assume that licensee mitigation
and recovery actions, including any
operator actions, are unavailable or
ineffective. The licensee must maintain
the analysis until the permanent

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cessation of operations and permanent
removal of fuel from the reactor vessel
as described under § 53.1070.
(b) Fitness for Duty. Each holder of an
OL or COL under this part must
develop, implement, and maintain a
fitness for duty program under 10 CFR
part 26.
(c) Access Authorization. Each holder
of an OL or COL under this part must
develop, implement, and maintain an
access authorization program under
§ 73.120 of this chapter if the criterion
in § 53.860(a)(2)(i) is satisfied, or the
requirements in § 73.56 of this chapter
if the criterion is not satisfied.
(d) Cybersecurity. Each holder of an
OL or COL under this part must
develop, implement, and maintain a
cybersecurity program under §§ 73.54 or
73.110 of this chapter.
(e) Information Security. Each holder
of an OL or COL under this part must
develop, implement, and maintain an
information protection system under
§§ 73.21, 73.22, and 73.23 of this
chapter, as applicable.
§ 53.865

Quality assurance.

Each holder of an OL or COL under
this part must develop, implement, and
maintain a quality assurance program in
accordance with appendix B of part 50
of this chapter. A written quality
assurance program manual must be
developed and used to guide the
conduct of the program in accordance
with generally accepted consensus
codes and standards that have been
endorsed or otherwise found acceptable
by the NRC.
§ 53.870

Integrity assessment programs.

Each holder of an OL or COL under
this part must develop, implement, and
maintain an integrity assessment
program to monitor, evaluate, and
manage—
(a) The effects of plant aging on SR
and NSRSS SSCs. The program may
refer to surveillances, tests, and
inspections conducted for specific SSCs
in accordance with other requirements
in this part or conducted in accordance
with applicable consensus codes and
standards endorsed or otherwise found
acceptable by the NRC;
(b) Cyclic or transient load limits to
ensure that SR and NSRSS SSCs are
maintained within the applicable design
limits; and
(c) Degradation mechanisms related to
chemical interactions, operating
temperatures, effects of irradiation, and
other environmental factors to ensure
that the capabilities, availability, and
reliability of SR and NSRSS SSCs
demonstrate compliance with the

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§ 53.880 Inservice inspection and
inservice testing.

functional design criteria of §§ 53.410
and 53.420.

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§ 53.875

Fire protection.

(a)(1) Each holder of an OL or COL
under this part must have a fire
protection plan that describes the
overall fire protection program for the
facility; identifies the various positions
within the licensee’s organization that
are responsible for the program; states
the authorities that are delegated to each
of these positions to implement those
responsibilities; and outlines the plans
for fire protection, fire detection and
suppression capability; and limitation of
fire damage.
(2) The fire protection plan must also
describe specific features necessary to
implement the program described in
paragraph (a)(1) of this section such as
the following: administrative controls
and personnel requirements for fire
prevention and manual fire suppression
activities; automatic and manually
operated fire detection and suppression
systems; and the means to limit fire
damage to SSCs so that the capability to
demonstrate compliance with the
requirements of § 53.210 is ensured.
(b)(1) Each holder of an OL or COL
under this part must develop a
performance-based or deterministic fire
protection program that demonstrates
compliance with the safety criteria
outlined in §§ 53.210 and 53.220,
related safety functions outlined in
§ 53.230, and defense in depth as
outlined in § 53.250 with specific fire
protection measures related to fire
prevention, fire detection, and fire
suppression.
(2) The fire protection program must
comply with the following:
(i) Safety-related and NSRSS SSCs
must be designed, located, and
maintained to minimize, consistent with
other safety requirements, the
probability and effect of fires and
explosions.
(ii) Noncombustible and fire-resistant
materials must be used wherever
practical throughout the facility,
particularly in locations with SR and
NSRSS SSCs.
(iii) Fire detection and fire
suppression systems of appropriate
capacity and capability must be
provided and designed and maintained
to minimize the adverse effects of fires
on SR and NSRSS SSCs.
(iv) Fire suppression systems must be
designed and maintained to ensure that
their rupture or inadvertent operation
does not significantly impair the ability
of SR and NSRSS SSCs to perform their
safety functions to satisfy § 53.230.

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(a) Each holder of an OL or COL
under this part must develop,
implement, and maintain a program for
inservice inspection (ISI) and inservice
testing (IST) prior to receiving an OL or
COL. The ISI/IST programs must,
wherever applicable, be in accordance
with generally accepted consensus
codes and standards that have been
endorsed or otherwise found acceptable
by the NRC. The ISI/IST program must
include all inspections and tests
required by the codes and standards
used in the design and be supplemented
by risk insights that identify the most
important SSCs to plant safety. The
types of testing and inspections and
their frequency should be informed by
risk insights to maintain the reliability
and performance of SSCs consistent
with the associated design and analyses
activities involving those SSCs. Risk
insights must also be used to determine
when to conduct the inspections and
tests (e.g., full power, shutdown,
refueling) to minimize risk to the plant
workers and the public. The ISI/IST
program must be documented in a
written manual and managed by
qualified personnel reporting to the
Plant Manager.
(b) Prior to plant operation, baseline
inspections and testing must be
performed using the same techniques as
will be used for future inspections and
testing. The results of these inspections
and testing must be used as benchmarks
for evaluating the results of future
inspections and testing. Sufficient room
and support must be provided to
accommodate the personnel, ISI/IST
equipment, and shielding necessary to
perform the inspections and testing.
Acceptance criteria for determining
whether corrective action is needed
must be developed (or taken from the
codes and standards used in the design)
for evaluating the results of the
inspections and testing. The results of
the inspections and testing must be
provided to the Plant Manager who is
responsible for determining what, if
any, corrective action is needed and
when it should be taken. The ISI/IST
results and corrective actions must be
documented and the documentation
retained for the life of the plant.
§ 53.910

Procedures and guidelines.

(a) Each holder of an OL or COL
under this part must have a program for
developing, implementing, and
maintaining an integrated set of
procedures, guidelines, and related
supporting activities to support normal
operations and respond to possible
unplanned events.

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(b) The program required by
paragraph (a) of this section must
include but is not limited to
development, implementation,
maintenance, and supporting activities
of procedures and guidelines for the
following:
(1) Plant operations;
(2) Maintenance activities under
§ 53.715;
(3) Program requirements under this
subpart F of this part;
(4) Emergency operating procedures,
if developed to address the role of
human actions in responding to LBEs;
(5) Accident management guidelines,
if developed to address the role of
human actions in responding to LBEs;
(6) Procedures for each area in which
licensed special nuclear material is
handled, used, or stored to protect
personnel upon the sounding of a
criticality alarm required by
§ 53.440(m); and
(7) Procedures that describe how the
licensee will address the following areas
if the licensee is notified of a potential
aircraft threat:
(i) Verification of the authenticity of
threat notifications;
(ii) Maintenance of continuous
communication with threat notification
sources;
(iii) Contacting all onsite personnel
and applicable offsite response
organizations;
(iv) Onsite actions necessary to
enhance the capability of the facility to
mitigate the consequences of an aircraft
impact;
(v) Measures to reduce visual
discrimination of the site relative to its
surroundings or individual buildings
within the protected area;
(vi) Dispersal of equipment and
personnel, as well as rapid entry into
site protected areas for essential onsite
personnel and offsite responders who
are necessary to mitigate the event; and
(vii) Recall of site personnel.
Subpart G—Decommissioning
Requirements
§ 53.1000

Scope and purpose.

This subpart defines the requirements
related to decommissioning for
applicants for, or holders of, an
operating license (OL) or combined
license (COL). The requirements related
to maintaining financial assurance for
decommissioning are in §§ 53.1010
through 53.1060. The requirements for
transitioning from operations to
decommissioning and for the release of
property and termination of the license
are in §§ 53.1070 through 53.1080.

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§ 53.1010 Financial assurance for
decommissioning.

(a) This section establishes
requirements for indicating to the U.S.
Nuclear Regulatory Commission (NRC)
how an applicant for or holder of an OL
or COL under this part will provide
reasonable assurance that funds will be
available for the decommissioning
process. Reasonable assurance consists
of a series of steps as provided in
paragraph (b) of this section and
§§ 53.1020, 53.1030 and 53.1040.
Funding for the decommissioning of
commercial nuclear plants may also be
subject to the regulation of Federal or
State government agencies (e.g., Federal
Energy Regulatory Commission (FERC)
and State Public Utility Commissions)
that have jurisdiction over rate
regulation. The requirements of this
subpart, in particular § 53.1020, are in
addition to, and not a substitution for,
other requirements, and are not
intended to be used by themselves or by
other agencies to establish rates.
(b) Each applicant for an OL or COL
under this part must prepare a plan and
an associated decommissioning report
that ensures and documents that
adequate funding will be available to
decommission the facility. Each holder
of an OL or COL must implement and
maintain the plan.
(1)(i) Before the Commission issues an
OL under this part, the applicant must
update the decommissioning report to
certify that it has provided financial
assurance for decommissioning in the
amount proposed in the application and
approved by the NRC under § 53.1020.
(ii) No later than 30 days after the
Commission issues the notice of
intended operation under § 53.1452 for
a COL under this part, the licensee must
update the decommissioning report to
certify that it has provided financial
assurance for decommissioning in the
amount proposed in the application and
approved by the NRC under § 53.1020.
(2) The amount of financial assurance
for decommissioning to be provided
must be based on a site-specific cost
estimate for decommissioning the
facility under § 53.1020.
(3) The amount of financial assurance
for decommissioning to be provided
must be adjusted annually using a rate
at least equal to that stated in § 53.1030.
(4) The amount of financial assurance
for decommissioning to be provided
must be covered by one or more of the
methods described in § 53.1040 as
acceptable to the NRC. A copy of the
financial instrument obtained to satisfy
the requirements of § 53.1040 must be
submitted to the NRC as part of the
application for an OL under this part;
however, an applicant for or holder of

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a COL need not obtain such financial
instrument or submit a copy to the
Commission except as provided in
§ 53.1060(b).

§ 53.1040 Methods for providing financial
assurance for decommissioning.

Financial assurance for
decommissioning is to be provided by
the following methods.
§ 53.1020 Cost estimates for
(a) Prepayment. Prepayment is the
decommissioning.
deposit made preceding the start of
operation or the transfer of a license
Cost estimates for decommissioning
under § 53.1570 into an account
must be site-specific. Site-specific
segregated from licensee assets and
decommissioning cost estimates (DCEs)
must account for the engineering, labor, outside the administrative control of the
equipment, transportation, disposal, and licensee and its subsidiaries or affiliates
of cash or liquid assets such that the
related charges needed to support
amount of funds would be sufficient to
termination of the license. They must
pay decommissioning costs. Prepayment
include the costs for decontaminating
may be in the form of a trust, escrow
structures, systems, and components
account, or Government fund with
and the site environs; removal of
contaminated components and materials payment by certificate of deposit,
deposit of government or other
from the plant and the site environs;
securities, or other method acceptable to
disposal of removed components and
the NRC. This trust, escrow account,
materials in appropriate facilities; and
Government fund, or other type of
any other activities supporting the
agreement must be established in
release of the property and termination
writing and maintained at all times in
of the license. They must also address
the United States with an entity that is
the approach to annual adjustments
an appropriate State or Federal
required by § 53.1030. Finally, siteGovernment agency, or an entity whose
specific DCEs must include plans for
operations in which the prepayment
adjusting levels of funds assured for
deposit is managed are regulated and
decommissioning to demonstrate that a
examined by a Federal or State agency.
reasonable level of assurance will be
A licensee that has prepaid funds based
provided that funds will be available
on a site-specific cost estimate under
when needed to cover the cost of
§ 53.1020 may take credit for projected
decommissioning.
earnings on the prepaid
§ 53.1030 Annual adjustments to cost
decommissioning trust funds, using up
estimates for decommissioning.
to a 2 percent annual real rate of return
through the time of termination of the
Each holder of an OL or COL under
license. A licensee may use a credit of
this part must annually adjust the cost
greater than 2 percent if the licensee’s
estimate for decommissioning to
rate-setting authority has specifically
account for escalation in labor, energy,
authorized a higher rate. Actual
and waste burial costs. Licensees may
earnings on existing funds may be used
elect to use either a site-specific
to calculate future fund needs.
adjustment factor, approved as part of
(b) External sinking fund. An external
the plan and associated
sinking fund is a fund established and
decommissioning report required by
maintained by setting funds aside
§ 53.1010, in paragraph (a) of this
periodically in an account segregated
section or the generic adjustment factor
from licensee assets and outside the
in paragraph (b) of this section.
administrative control of the licensee
(a) A site-specific adjustment factor
and its subsidiaries or affiliates in
must address the estimated
which the total amount of funds would
contributions and escalation of costs for
be sufficient to pay decommissioning
the following aspects of
costs. An external sinking fund may be
decommissioning:
in the form of a trust, escrow account,
(1) Labor, materials, and services;
or Government fund, with payment by
(2) Energy and waste transportation;
certificate of deposit, deposit of
and
Government or other securities, or other
(3) Radioactive waste burial or other
method acceptable to the NRC. This
disposition.
trust, escrow account, Government
(b) A generic adjustment factor must
fund, or other type of agreement must be
be at least equal to 0.65 L + 0.13 E + 0.22 established in writing and maintained at
B, where L and E are escalation factors
all times in the United States with an
for labor and energy, respectively, and
entity that is an appropriate State or
are to be taken from regional data of
Federal Government agency, or an entity
U.S. Department of Labor Bureau of
whose operations in which the external
Labor Statistics and B is an escalation
sinking fund is managed are regulated
factor for waste burial and is to be taken and examined by a Federal or State
from NRC report NUREG–1307, ‘‘Report agency. A licensee that has collected
on Waste Burial Charges.’’
funds based on a site-specific cost

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estimate under § 53.1020 may take
credit for projected earnings on the
external sinking funds using up to a 2
percent annual real rate of return from
the time of future funds’ collection
through the time of termination of the
license. A licensee may use a credit of
greater than 2 percent if the licensee’s
rate-setting authority has specifically
authorized a higher rate. Actual
earnings on existing funds may be used
to calculate future fund needs. A
licensee whose rates for
decommissioning costs cover only a
portion of these costs may make use of
this method only for the portion of these
costs that are collected in one of the
manners described in this paragraph.
This method may be used as the
exclusive mechanism relied upon for
providing financial assurance for
decommissioning in the following
circumstances:
(1) By a licensee that recovers, either
directly or indirectly, the estimated total
cost of decommissioning through rates
established by ‘‘cost of service’’ or
similar ratemaking regulation. Public
utility districts, municipalities, rural
electric cooperatives, and State and
Federal agencies, including associations
of any of the foregoing, that establish
their own rates and are able to recover
their cost of service allocable to
decommissioning, are deemed to satisfy
this condition.
(2) By a licensee whose source of
revenues for its external sinking fund is
a ‘‘non-bypassable charge,’’ the total
amount of which will provide funds
estimated to be needed for
decommissioning pursuant to
§§ 53.1020, 53.1060, or 53.1575.
(c) A surety method, insurance, or
other guarantee method. (1) These
methods guarantee that
decommissioning costs will be paid. A
surety method may be in the form of a
surety bond, or letter of credit. Any
surety method or insurance used to
provide financial assurance for
decommissioning must contain the
following conditions:
(i) The surety method or insurance
must be open-ended, or, if written for a
specified term, such as 5 years, must be
renewed automatically, unless 90 days
or more prior to the renewal day the
issuer notifies the NRC, the beneficiary,
and the licensee of its intention not to
renew. The surety or insurance must
also provide that the full-face amount be
paid to the beneficiary automatically
prior to the expiration without proof of
forfeiture if the licensee fails to provide
a replacement acceptable to the NRC
within 30 days after receipt of
notification of cancellation.

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(ii) The surety or insurance must be
payable to a trust established for
decommissioning costs. The trustee and
trust must be acceptable to the NRC. An
acceptable trustee includes an
appropriate State or Federal
Government agency or an entity that has
the authority to act as a trustee and
whose trust operations are regulated and
examined by a Federal or State agency.
(2) A parent company guarantee of
funds for decommissioning costs based
on a financial test may be used if the
guarantee and test are as contained in
appendix A to 10 CFR part 30.
(3) For commercial companies that
issue bonds, a guarantee of funds by the
applicant or licensee for
decommissioning costs based on a
financial test may be used if the
guarantee and test are as contained in
appendix C to 10 CFR part 30. For
commercial companies that do not issue
bonds, a guarantee of funds by the
applicant or licensee for
decommissioning costs may be used if
the guarantee and test are as contained
in appendix D to 10 CFR part 30. A
guarantee by the applicant or licensee
may not be used in any situation in
which the applicant or licensee has a
parent company holding majority
control of voting stock of the company.
(d) Funding method for Federal
licensees. For a Federal licensee, a
statement of intent containing a cost
estimate for decommissioning and
indicating that funds for
decommissioning will be obtained when
necessary.
(e) Contractual funding method.
Contractual obligation(s) on the part of
a licensee’s customer(s), the total
amount of which over the duration of
the contract(s) will provide the
licensee’s total share of uncollected
funds estimated to be needed for
decommissioning pursuant to
§§ 53.1020, 53.1060, or 53.1575. To be
acceptable to the NRC as a method of
decommissioning funding assurance,
the terms of the contract(s) must include
provisions that the buyer(s) of electricity
or other products will pay for the
decommissioning obligations specified
in the contract(s), notwithstanding the
operational status either of the licensed
plant to which the contract(s) pertains
or force majeure provisions. All
proceeds from the contract(s) for
decommissioning funding will be
deposited to the external sinking fund.
The NRC reserves the right to evaluate
the terms of any contract(s) and the
financial qualifications of the
contracting entity or entities offered as
assurance for decommissioning funding.
(f) Other funding mechanisms. Any
other mechanism, or combination of

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mechanisms, that provides, as
determined by the NRC upon its
evaluation of the specific circumstances
of each licensee submittal, assurance of
decommissioning funding equivalent to
that provided by the mechanisms
specified in paragraphs (a) through (e) of
this section. Licensees who do not have
sources of funding described in
paragraph (b) of this section may use an
external sinking fund in combination
with a guarantee mechanism, as
specified in paragraph (c) of this
section, provided that the total amount
of funds estimated to be necessary for
decommissioning is assured.
§ 53.1045 Limitations on the use of
decommissioning trust funds.

(a)(1) Decommissioning trust funds
may be used by licensees if—
(i) The withdrawals are for expenses
for decommissioning activities
consistent with the definition of
decommission or decommissioning in
§ 53.020;
(ii) The expenditure would not reduce
the value of the decommissioning trust
below an amount necessary to place and
maintain the reactor in a safe storage
condition if unforeseen conditions or
expenses arise; and
(iii) The withdrawals would not
inhibit the ability of the licensee to
complete funding of any shortfalls in
the decommissioning trust needed to
ensure the availability of funds to
ultimately release the site and terminate
the license.
(2) Initially, 3 percent of the amount
determined in accordance with
§ 53.1020 may be used for
decommissioning planning. For
licensees that have submitted the
certifications required under § 53.1070
and commencing 90 days after the NRC
has received the post-shutdown
decommissioning activities report
(PSDAR) required by § 53.1060, an
additional 20 percent may be used. An
updated site-specific DCE must be
submitted to the NRC prior to the
licensee using any funding in excess of
these amounts.
(b) Licensees that are not ‘‘electric
utilities’’ as defined in § 53.020 that use
prepayment or an external sinking fund
to provide financial assurance must
provide in the terms of the arrangements
governing the trust, escrow account, or
Government fund, used to segregate and
manage the funds that—
(1) The trustee, manager, investment
advisor, or other person directing
investment of the funds—
(i) Is prohibited from investing the
funds in securities or other obligations
of the licensee or any other owner or
operator of any commercial nuclear

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plant or their affiliates, subsidiaries,
successors or assigns, or in a mutual
fund in which at least 50 percent of the
fund is invested in the securities of a
licensee or parent company whose
subsidiary is an owner or operator of a
foreign or domestic commercial nuclear
plant. However, the funds may be
invested in securities tied to market
indices or other non-nuclear sector
collective, commingled, or mutual
funds, provided that no more than 10
percent of trust assets may be indirectly
invested in securities of any entity
owning or operating one or more
commercial nuclear plants.
(ii) Is obligated at all times to adhere
to a standard of care set forth in the
trust, which either shall be the standard
of care, whether in investing or
otherwise, required by State or Federal
law or one or more State or Federal
regulatory agencies with jurisdiction
over the trust funds, or, in the absence
of any such standard of care, whether in
investing or otherwise, that a prudent
investor would use in the same
circumstances. The term ‘‘prudent
investor,’’ shall have the same meaning
as set forth in FERC’s ‘‘Regulations
Governing Nuclear Plant
Decommissioning Trust Funds’’ at 18
CFR 35.32(a)(3), or any successor
regulation.
(2) The licensee, its affiliates, and its
subsidiaries are prohibited from being
engaged as investment manager for the
funds or from giving day-to-day
management direction of the funds’
investments or direction on individual
investments by the funds, except in the
case of passive fund management of
trust funds where management is
limited to investments tracking market
indices.
(3) The trust, escrow account,
Government fund, or other account used
to segregate and manage the funds may
not be amended in any material respect
without written notification to the
Director, Office of Nuclear Reactor
Regulation, or Director, Office of
Nuclear Material Safety and Safeguards,
as applicable, at least 30 working days
before the proposed effective date of the
amendment. The licensee must provide
the text of the proposed amendment and
a statement of the reason for the
proposed amendment. The trust, escrow
account, Government fund, or other
account may not be amended if the
person responsible for managing the
trust, escrow account, Government
fund, or other account receives written
notice of objection from the Director,
Office of Nuclear Reactor Regulation, or
Director, Office of Nuclear Material
Safety and Safeguards, as applicable,
within the notice period.

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(4) Except for withdrawals being
made under paragraph (a) of this section
or for payments of ordinary
administrative costs (including taxes)
and other incidental expenses of the
fund (including legal, accounting,
actuarial, and trustee expenses) in
connection with the operation of the
fund, no disbursement or payment may
be made from the trust, escrow account,
Government fund, or other account used
to segregate and manage the funds until
written notice of the intention to make
a disbursement or payment has been
given to the Director, Office of Nuclear
Reactor Regulation, or Director, Office of
Nuclear Material Safety and Safeguards,
as applicable, at least 30 working days
before the date of the intended
disbursement or payment. The
disbursement or payment from the trust,
escrow account, Government fund or
other account may be made following
the 30 working day notice period if the
person responsible for managing the
trust, escrow account, Government
fund, or other account does not receive
written notice of objection from the
Director, Office of Nuclear Reactor
Regulation, or Director, Office of
Nuclear Material Safety and Safeguards,
as applicable, within the notice period.
Disbursements or payments from the
trust, escrow account, Government
fund, or other account used to segregate
and manage the funds, other than for
payment of ordinary administrative
costs (including taxes) and other
incidental expenses of the fund
(including legal, accounting, actuarial,
and trustee expenses) in connection
with the operation of the fund, are
restricted to decommissioning expenses
or transfer to another financial
assurance method acceptable under
§ 53.1040 until final decommissioning
has been completed. After
decommissioning has begun and
withdrawals from the decommissioning
fund are made under paragraph (a) of
this section, no further notification need
be made to the NRC.
(c) Licensees that are ‘‘electric
utilities’’ under § 53.020 that use
prepayment or an external sinking fund
to provide financial assurance must
include a provision in the terms of the
trust, escrow account, Government
fund, or other account used to segregate
and manage funds that except for
withdrawals being made under
paragraph (a) of this section or for
payments of ordinary administrative
costs (including taxes) and other
incidental expenses of the fund
(including legal, accounting, actuarial,
and trustee expenses) in connection
with the operation of the fund, no

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disbursement or payment may be made
from the trust, escrow account,
Government fund, or other account used
to segregate and manage the funds until
written notice of the intention to make
a disbursement or payment has been
given the Director, Office of Nuclear
Reactor Regulation, or Director, Office of
Nuclear Material Safety and Safeguards,
as applicable, at least 30 working days
before the date of the intended
disbursement or payment. The
disbursement or payment from the trust,
escrow account, Government fund or
other account may be made following
the 30 working day notice period if the
person responsible for managing the
trust, escrow account, Government
fund, or other account does not receive
written notice of objection from the
Director, Office of Nuclear Reactor
Regulation, or Director, Office of
Nuclear Material Safety and Safeguards,
as applicable, within the notice period.
Disbursements or payments from the
trust, escrow account, Government
fund, or other account used to segregate
and manage the funds, other than for
payment of ordinary administrative
costs (including taxes) and other
incidental expenses of the fund
(including legal, accounting, actuarial,
and trustee expenses) in connection
with the operation of the fund, are
restricted to decommissioning expenses
or transfer to another financial
assurance method acceptable under
§ 53.1040 until final decommissioning
has been completed. After
decommissioning has begun and
withdrawals from the decommissioning
fund are made under paragraph (a) of
this section, no further notification need
be made to the NRC.
(d) A licensee that is not an ‘‘electric
utility’’ under § 53.020 and using a
surety method, insurance, or other
guarantee method to provide financial
assurance must provide that the trust
established for decommissioning costs
to which the surety or insurance is
payable contains in its terms the
requirements in § 53.1045(b)(1) through
(4).
§ 53.1050

NRC oversight.

The NRC reserves the right to take the
following steps in order to ensure a
licensee’s adequate accumulation of
decommissioning funds: review, as
needed, the rate of accumulation of
decommissioning funds and, either
independently or in cooperation with
FERC and the licensee’s State Public
Utility Commission, take additional
actions as appropriate on a case-by-case
basis, including modification of a
licensee’s schedule for the accumulation
of decommissioning funds.

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§ 53.1060 Reporting and recordkeeping
requirements.

(a) Each holder of an OL under this
part or holder of a COL under this part
after the date that the Commission has
made the finding under § 53.1452(g)
must report, at least once every 2 years,
by March 31, on the status of its
certification of decommissioning
funding for each commercial nuclear
reactor or part of a commercial nuclear
reactor that it owns. The information in
this report must include, at a minimum,
the amount of decommissioning funds
estimated to be required under
§§ 53.1020 and 53.1030; the amount of
decommissioning funds accumulated to
the end of the calendar year preceding
the date of the report; a schedule of the
annual amounts remaining to be
collected; the assumptions used
regarding rates of escalation in
decommissioning costs, rates of
earnings on decommissioning funds,
and rates of other factors used in
funding projections; any contracts upon
which the licensee is relying under
§ 53.1040(e); any modifications
occurring to a licensee’s method of
providing financial assurance since the
last submitted report; and any material
changes to trust agreements. If any of
the preceding items is not applicable,
the licensee should so state in its report.
Any licensee for a plant that is within
5 years of the projected end of its
operation, or where conditions have
changed such that it will close within 5
years (before the end of its licensed life),
or that has already closed (before the
end of its licensed life), or that is
involved in a merger or an acquisition
must submit this report annually.
(b) Each holder of a COL under this
part must, 2 years before and 1 year
before the scheduled date for initial
loading of fuel (or, for a fueled
manufactured reactor, 2 years before
and 1 year before the scheduled date for
initiating the physical removal of any
one of the independent physical
mechanisms to prevent criticality
required under § 53.620(d)(1)), submit a
report to the NRC containing a
certification updating the DCEs and a
copy of the financial instrument to be
used to satisfy § 53.1040. No later than
30 days after the Commission publishes
notice in the Federal Register under
§ 53.1452(a), the licensee must submit
an updated decommissioning report
required under § 53.1010(b)(1)(ii),
including a copy of the financial
instrument obtained to satisfy § 53.1040.
(c) Each licensee must keep records of
information important to the safe and
effective decommissioning of the facility
in an identified location until the
license is terminated by the

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Commission. If records of relevant
information are kept for other purposes,
reference to these records and their
locations may be used. Information the
Commission considers important to
decommissioning consists of—
(1) Records of spills or other unusual
occurrences involving the spread of
contamination in and around the
facility, equipment, or site. These
records may be limited to instances
when significant contamination remains
after any cleanup procedures or when
there is reasonable likelihood that
contaminants may have spread to
inaccessible areas as in the case of
possible seepage into porous materials
such as concrete. These records must
include any known information on
identification of involved nuclides,
quantities, forms, and concentrations.
(2) As-built drawings and
modifications of structures and
equipment in restricted areas where
radioactive materials are used and/or
stored and of locations of possible
inaccessible contamination such as
buried pipes that may be subject to
contamination. If required drawings are
referenced, each relevant document
need not be indexed individually. If
drawings are not available, the licensee
must substitute appropriate records of
available information concerning these
areas and locations.
(3) Records of the cost estimate
performed for the decommissioning
funding plan or of the amount certified
for decommissioning, and records of the
funding method used for assuring funds
if either a funding plan or certification
is used.
(4) Records of—
(i) The licensed site area, as originally
licensed and any revisions, which must
include a site map and any acquisition
or use of property outside the originally
licensed site area for the purpose of
receiving, possessing, or using licensed
materials;
(ii) The licensed activities carried out
on the acquired or used property; and
(iii) The release and final disposition
of any property recorded in paragraph
(c)(4)(i) of this section, the historical site
assessment performed for the release,
radiation surveys performed to support
release of the property, submittals to the
NRC made under § 53.1070, and the
methods employed to ensure that the
property met the radiological criteria of
subpart E of 10 CFR part 20 at the time
the property was released.
(d) Each holder of an OL or COL
under this part must at or about 5 years
prior to the projected end of operations
submit a preliminary DCE which
includes an up-to-date assessment of the

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major factors that could affect the cost
to decommission.
(e) Prior to or within 2 years following
permanent cessation of operations, the
licensee must submit a PSDAR to the
NRC, and a copy to the affected State(s).
The PSDAR must contain a description
of the planned decommissioning
activities along with a schedule for their
accomplishment, a discussion that
provides the reasons for concluding that
the environmental impacts associated
with site-specific decommissioning
activities will be bounded by
appropriate previously issued
environmental impact statements, and a
site-specific DCE, including the
projected cost of managing irradiated
fuel.
(f) For decommissioning activities
that delay completion of
decommissioning by including a period
of storage or surveillance, the licensee
must provide a means of adjusting cost
estimates and associated funding levels
over the storage or surveillance period.
(g) After submitting its site-specific
DCE required by paragraph (e) of this
section, and until the licensee has
completed its final radiation survey and
demonstrated that residual radioactivity
has been reduced to a level that permits
termination of its license, the licensee
must annually submit to the NRC, by
March 31, a financial assurance status
report. The report must include the
following information, current through
the end of the previous calendar year:
(1) The amount spent on
decommissioning, both cumulative and
over the previous calendar year, the
remaining balance of any
decommissioning funds, and the
amount provided by other financial
assurance methods being relied upon;
(2) An estimate of the costs to
complete decommissioning, reflecting
any difference between actual and
estimated costs for work performed
during the year, and the
decommissioning criteria upon which
the estimate is based;
(3) Any modifications occurring to a
licensee’s current method of providing
financial assurance since the last
submitted report; and
(4) Any material changes to trust
agreements or financial assurance
contracts.
(5) If the sum of the balance of any
remaining decommissioning funds, plus
earnings on such funds calculated at not
greater than a 2 percent real rate of
return, together with the amount
provided by other financial assurance
methods being relied upon, does not
cover the estimated cost to complete the
decommissioning, the financial
assurance status report must include

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additional financial assurance to cover
the estimated cost of completion.
(h) After submitting its site-specific
DCE required by paragraph (e) of this
section, the licensee must annually
submit to the NRC, by March 31, a
report on the status of its funding for
managing irradiated fuel. The report
must include the following information,
current through the end of the previous
calendar year:
(1) The amount of funds accumulated
to cover the cost of managing the
irradiated fuel;
(2) The projected cost of managing
irradiated fuel until title to the fuel and
possession of the fuel is transferred to
the Secretary of Energy; and
(3) If the funds accumulated do not
cover the projected cost, a plan to obtain
additional funds to cover the cost.

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§ 53.1070

Termination of license.

For each holder of an OL or COL
under this part—
(a)(1) When the licensee has
determined to permanently cease
operations the licensee must, within 30
days, submit a written certification to
the NRC, consistent with the
requirements of § 53.040(b)(8);
(2) When appropriate to support
decommissioning activities and the
eventual permanent removal of fuel
from the reactor vessel, the licensee
must develop defueled technical
specifications by reviewing the
operational technical specifications and
determining which specifications no
longer apply during decommissioning
and which ones should remain
applicable. The licensee must make the
appropriate submittals to the NRC in
accordance with § 53.1510 to request
changes to the technical specifications;
and
(3)(i) Once fuel has been permanently
removed from the reactor vessel, the
licensee must submit a written
certification to the NRC that meets the
requirements of § 53.040(b)(9); and
(ii) The licensee must establish and
maintain staffing consisting of certified
fuel handlers, as defined under § 53.020,
and other non-licensed personnel with
appropriate qualifications, and in
sufficient numbers, to ensure support
for facility operations and radiological
control activities, as required by the
facility defueled technical
specifications. These personnel must be
subject to the training requirements of
§ 53.830.
(b) Upon docketing of the
certifications for permanent cessation of
operations and permanent removal of
fuel from the reactor vessel, or when a
final legally effective order to
permanently cease operations has come

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into effect, the license issued under this
part no longer authorizes operation of
the reactor or emplacement or retention
of fuel into the reactor vessel.
(c) Decommissioning will be
completed within 60 years of permanent
cessation of operations. Completion of
decommissioning beyond 60 years will
be approved by the Commission only
when necessary to protect public health
and safety. Factors that will be
considered by the Commission in
evaluating an alternative that provides
for completion of decommissioning
beyond 60 years of permanent cessation
of operations include unavailability of
waste disposal capacity and other sitespecific factors affecting the licensee’s
capability to carry out
decommissioning, including presence of
other nuclear facilities at the site.
(d)(1) Prior to or within 2 years
following permanent cessation of
operations, the licensee must submit a
PSDAR and site-specific DCE in
accordance with § 53.1060(e).
(2) The NRC must notice receipt of the
PSDAR and make the PSDAR publicly
available and publish notice of its
availability for public comment in the
Federal Register. The NRC must also
schedule a public meeting readily
accessible to individuals in the vicinity
of the licensee’s facility. The NRC must
publish a notice in the Federal Register
and in a forum, such as local
newspapers, that is readily accessible to
individuals in the vicinity of the site,
announcing the date, time, and location
of the meeting, along with a brief
description of the purpose of the
meeting.
(e) Licensees must not perform any
major decommissioning activities, as
defined in § 53.020, until 90 days after
the NRC has received the licensee’s
PSDAR submittal and until
certifications of permanent cessation of
operations and permanent removal of
fuel from the reactor vessel, as required
under paragraph (a) of this section, have
been submitted.
(f) Licensees must not perform any
decommissioning activities, as defined
in § 53.020, that—
(1) Foreclose release of the site for
possible unrestricted use;
(2) Result in significant
environmental impacts not previously
reviewed; or
(3) Result in there no longer being
reasonable assurance that adequate
funds will be available for
decommissioning.
(g) In taking actions permitted under
§ 53.1540 following submittal of the
PSDAR, the licensee must notify the
NRC in writing, and send a copy to the
affected State(s), before performing any

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decommissioning activity inconsistent
with, or making any significant
schedule change from, those actions and
schedules described in the PSDAR,
including changes that increase the
decommissioning cost by more than 20
percent from the previously provided
DCE.
(h) Licensees may use
decommissioning trust funds consistent
with the limitations of § 53.1045(a).
Licensees must report on the status of
decommissioning trust funds consistent
with the requirements of § 53.1060.
(i) Licensees must submit an
application for termination of license in
accordance with § 53.1070. The
application for termination of license
must be accompanied or preceded by a
license termination plan to be submitted
for NRC approval.
(1) The license termination plan must
be a supplement to the Final Safety
Analysis Report or equivalent and must
be submitted at least 2 years before
termination of the license date.
(2) The license termination plan must
include—
(i) A site characterization;
(ii) Identification of remaining
dismantlement activities;
(iii) Plans for site remediation;
(iv) Detailed plans for the final
radiation survey;
(v) A description of the end use of the
site, if restricted;
(vi) An updated site-specific estimate
of remaining decommissioning costs;
(vii) A supplement to the
environmental report, pursuant to
§ 51.53 of this chapter, describing any
new information or significant
environmental change associated with
the licensee’s proposed termination
activities; and
(viii) Identification of parts, if any, of
the facility or site that were released for
use before approval of the license
termination plan.
(3) Following receipt of the license
termination plan, the NRC must make
the license termination plan publicly
available and publish notice of its
availability for public comment in the
Federal Register. The NRC must also
schedule a public meeting readily
accessible to individuals in the vicinity
of the licensee’s facility upon receipt of
the license termination plan. The NRC
must publish a notice in the Federal
Register and in a forum, such as local
newspapers, that is readily accessible to
individuals in the vicinity of the site,
announcing the date, time, and location
of the meeting, along with a brief
description of the purpose of the
meeting.
(j) If the license termination plan
demonstrates that the remainder of

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decommissioning activities will be
performed in accordance with the
regulations in this chapter, will not be
inimical to the common defense and
security or to the health and safety of
the public, and will not have a
significant effect on the quality of the
environment and after notice to
interested persons, the Commission will
approve the plan, by license
amendment, subject to such conditions
and limitations as it deems appropriate
and necessary and authorize
implementation of the license
termination plan.
(k) The Commission will terminate
the license if it determines that—
(1) The remaining dismantlement has
been performed in accordance with the
approved license termination plan, and
(2) The final radiation survey and
associated documentation, including an
assessment of dose contributions
associated with parts released for use
before approval of the license
termination plan, demonstrate that the
facility and site have met the criteria for
decommissioning in subpart E of 10
CFR part 20.

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§ 53.1075 Program requirements during
decommissioning.

(a) Licensees that have submitted the
certifications required under § 53.1070
must maintain a decommissioning fire
protection program to address the
potential for fires that could cause the
release or spread of radioactive
materials.
(1) The objectives of the
decommissioning fire protection
program are to
(i) Reasonably prevent these fires from
occurring;
(ii) Rapidly detect, control, and
extinguish those fires that do occur and
that could result in a radiological
hazard; and
(iii) Ensure that the risk of fireinduced radiological hazards to the
public, environment, and plant
personnel is minimized.
(2) The licensee must assess the
decommissioning fire protection
program on a regular basis. The licensee
must revise the decommissioning fire
protection program documentation as
appropriate throughout the various
stages of facility decommissioning.
(3) The licensee may make changes to
the decommissioning fire protection
program without NRC approval if these
changes do not reduce the effectiveness
of fire protection for structures, systems,
and components that could result in a
radiological hazard, taking into account
the decommissioning plant conditions
and activities.
(b) [Reserved]

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§ 53.1080 Release of part of a commercial
nuclear plant or site for unrestricted use.

(a) Prior written NRC approval is
required to release part of a commercial
nuclear plant or site for unrestricted use
at any time before receiving approval of
a license termination plan. Section
53.1060 specifies recordkeeping
requirements associated with partial
release. Holders of an OL or COL under
this part seeking NRC review and
approval must—
(1) Evaluate the effect of releasing the
property to ensure that—
(i) The dose to individual members of
the public does not exceed the limits
and standards of subpart D of 10 CFR
part 20;
(ii) There is no reduction in the
effectiveness of emergency planning or
physical security;
(iii) Effluent releases remain within
license conditions;
(iv) The environmental monitoring
program and offsite dose calculation
manual are revised to account for the
changes;
(v) The siting criteria of subpart D of
this part continue to be met; and
(vi) All other applicable statutory and
regulatory requirements continue to be
met.
(2) Perform a historical site
assessment of the part of the commercial
nuclear plant or site to be released; and
(3) Perform surveys adequate to
demonstrate compliance with the
radiological criteria for unrestricted use
specified in § 20.1402 of this chapter for
impacted areas.
(b) For release of non-impacted areas,
the licensee may submit a written
request for NRC review and approval of
the release if a license amendment is not
otherwise required. The request
submittal must include—
(1) The results of the evaluations
performed in accordance with
paragraphs (a)(1) and (a)(2) of this
section;
(2) A description of the part of the
commercial nuclear plant or site to be
released;
(3) The schedule for release of the
property;
(4) The results of the evaluations
performed in accordance with
§ 53.1540; and
(5) A discussion that provides the
reasons for concluding that the
environmental impacts associated with
the licensee’s proposed release of the
property will be bounded by
appropriate previously issued
environmental impact statements.
(c) After receiving a request from the
licensee for NRC approval of the release
of a non-impacted area, the NRC must—
(1) Determine whether the licensee
has adequately evaluated the effect of

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releasing the property as required by
paragraph (a)(1) of this section;
(2) Determine whether the licensee’s
classification of any release areas as
non- impacted is adequately justified;
and
(3) If determining that the licensee’s
submittal is adequate, inform the
licensee in writing that the release is
approved.
(d) For release of impacted areas, the
licensee must submit an application for
amendment of its license for the release
of the property. The application must
include—
(1) The information specified in
paragraphs (b)(1) through (b)(3) of this
section;
(2) The methods used for and results
obtained from the radiation surveys
required to demonstrate compliance
with the radiological criteria for
unrestricted use specified in § 20.1402;
and
(3) A supplement to the
environmental report, under § 51.53 of
this chapter, describing any new
information or significant
environmental change associated with
the licensee’s proposed release of the
property.
(e) After receiving a license
amendment application from the
licensee for the release of an impacted
area, the NRC must—
(1) Determine whether the licensee
has adequately evaluated the effect of
releasing the property as required by
paragraph (a)(1) of this section;
(2) Determine whether the licensee’s
classification of any release areas as
non-impacted is adequately justified;
(3) Determine whether the licensee’s
radiation survey for an impacted area is
adequate; and
(4) If determining that the licensee’s
submittal is adequate, approve the
licensee’s amendment application.
(f) The NRC must publish notice
receipt of the release approval request or
license amendment application in the
Federal Register and make the approval
request or license amendment
application available for public
comment. Before acting on an approval
request or license amendment
application submitted in accordance
with this section, the NRC must conduct
a public meeting readily accessible to
individuals in the vicinity of the
licensee’s facility for the purpose of
obtaining public comments on the
proposed release of part of the
commercial nuclear plant or site. The
NRC must publish a document in the
Federal Register and in a forum, such
as local newspapers, which is readily
accessible to individuals in the vicinity
of the site, announcing the date, time,

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and location of the meeting, along with
a brief description of the purpose of the
meeting.
Subpart H—Licenses, Certifications,
and Approvals

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§ 53.1100 Filing of application for licenses,
certifications, or approvals; oath or
affirmation.

(a) Serving of applications. (1) Each
filing of an application for a standard
design approval, standard design
certification, or license under this part,
and any amendments to the
applications, must be submitted to the
U.S. Nuclear Regulatory Commission
(NRC) under § 53.040, as applicable.
(2) Each applicant for a construction
permit (CP), early site permit, combined
license (COL), or manufacturing license
(ML) under this part must, upon
notification by the presiding officer
designated to conduct the public
hearing required by the Atomic Energy
Act of 1954, as amended, (the Act)
update the application and serve the
updated copies of the application or
parts of it, eliminating all superseded
information, together with an index of
the updated application, as directed by
presiding officer. Any subsequent
amendment to the application must be
served on those served copies of the
application and must be submitted to
the NRC as specified in § 53.040, as
applicable.
(3) The applicant must make a copy
of the updated application available at
the public hearing for the use of any
other parties to the proceeding and must
certify that the updated copies of the
application contain the current contents
of the application submitted in
accordance with the requirements under
this part.
(4) At the time of filing an
application, the Commission will make
available at the NRC website, http://
www.nrc.gov, a copy of the application,
subsequent amendments, and other
records pertinent to the matter that is
the subject of the application for public
inspection and copying.
(5) The serving of copies required by
this section must not occur until the
application has been docketed under
§ 2.101(a) of this chapter. Copies must
be submitted to the Commission, as
specified in § 53.040, as applicable, to
enable the Director, Office of Nuclear
Reactor Regulation to determine
whether the application is sufficiently
complete to permit docketing.
(b) Oath or affirmation. Each
application for a standard design
approval, standard design certification,
or license, including, whenever
appropriate, a CP or early site permit, or
amendment of it, and each amendment

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of each application must be executed in
a signed original by the applicant or
duly authorized officer thereof under
oath or affirmation.
(c) [Reserved]
(d) [Reserved]
(e) Filing fees. Each application for a
standard design approval, standard
design certification, or commercial
nuclear plant license under this part,
including, whenever appropriate, a CP,
COL, operating license (OL), ML, or
early site permit, other than a license
exempted from 10 CFR part 170, must
be accompanied by the fee prescribed in
10 CFR part 170. No fee will be required
to accompany an application for
renewal, amendment, or termination of
a CP, OL, COL, or ML, except as
provided in § 170.21 of this chapter.
(f) Environmental report. An
application for a CP, OL, early site
permit, design certification, COL, or ML
for a commercial nuclear plant must be
accompanied by an environmental
report required under subpart A of 10
CFR part 51.
§ 53.1101

Requirement for license.

Except as provided in § 53.1120, no
person within the United States may
transfer or receive in interstate
commerce, manufacture, produce,
transfer, acquire, possess, or use any
utilization facility except as authorized
by a license issued by the Commission.
§ 53.1103
licenses.

Combining applications and

(a) An applicant may combine several
applications in one application for
different kinds of licenses under the
regulations in this chapter.
(b) The Commission may combine in
a single license the activities of an
applicant which would otherwise be
licensed separately.
§ 53.1106

Elimination of repetition.

An applicant may incorporate by
reference in its application information
contained in previous applications,
statements, or reports filed with the
Commission, provided, however, that
such references are clear and specific.
§ 53.1109 Contents of applications;
general information.

Each application must include, unless
otherwise indicated in this subpart—
(a) Name of applicant;
(b) Address of applicant;
(c) Description of business or
occupation of applicant;
(d)(1) If applicant is an individual, the
citizenship of applicant;
(2) If applicant is a partnership, the
name, citizenship and address of each
partner and the principal location where
the partnership does business;

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(3) If applicant is a corporation or an
unincorporated association, the
following information:
(i) The State where it is incorporated
or organized and the principal location
where it does business;
(ii) The names, addresses and
citizenship of its directors and of its
principal officers; and
(iii) Whether it is owned, controlled,
or dominated by an alien, a foreign
corporation, or foreign government, and
if so, give details; or
(4) If the applicant is acting as agent
or representative of another person in
filing the application, identify the
principal and furnish information
required under this paragraph with
respect to such principal;
(e) The class and type of license
applied for, the use to which the facility
will be put, the period of time for which
the license is sought, and a list of other
licenses, except operator’s licenses,
issued or applied for in connection with
the proposed facility;
(f) [Reserved]
(g)(1) Except as provided in paragraph
(g)(2) of this section, if the application
is for an OL or COL for a commercial
nuclear plant, or if the application is for
an early site permit for a commercial
nuclear plant and contains plans for
coping with emergencies under
§ 53.1146(b)(2)(ii), the applicant must
submit the radiological emergency
response plans of State, local, and
participating Tribal governmental
entities in the United States that are
wholly or partially within the plume
exposure pathway emergency planning
zone (EPZ),1 and the plans of State
governments wholly or partially within
the ingestion pathway EPZ.2 If the
application is for an early site permit
that, under § 53.1146(b)(2)(i), proposes
major features of the emergency plans
describing the EPZs, then the
descriptions of the EPZs must meet the
requirements of this paragraph.
Generally, the plume exposure pathway
EPZ for a commercial nuclear plant
must consist of an area about 10 miles
(16 km) in radius and the ingestion
pathway EPZ must consist of an area
about 50 miles (80 km) in radius. The
exact size and configuration of the EPZs
surrounding a particular commercial
nuclear plant must be determined in
relation to the local emergency response
needs and capabilities as they are
affected by such conditions as
demography, topography, land
characteristics, access routes, and
jurisdictional boundaries. The size of
the EPZs also may be determined on a
case-by-case basis for gas-cooled
reactors and for reactors with an
authorized power level less than 250

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megawatt thermal. The plans for the
ingestion pathway must focus on such
actions as are appropriate to protect the
food ingestion pathway.
(2) Applicants for commercial nuclear
plants consisting of either small
modular reactors or non-light-water
reactors complying with § 50.160 of this
chapter who apply for a CP, an OL, a
COL, or an early site permit under this
part must submit as part of the
application the analysis used to
determine whether the criteria in
§ 53.1109(g)(2)(i)(A) and (B) are met
and, if they are met, the size of the
plume exposure pathway EPZ.
(i) The plume exposure pathway EPZ
is the area within which:
(A) Public dose, as defined in
§ 20.1003 of this chapter, is projected to
exceed 10 millisieverts (1 rem) total
effective dose equivalent over 96 hours
from the release of radioactive materials
from the facility considering accident
likelihood and source term, timing of
the accident sequence, and meteorology;
and
(B) Pre-determined, prompt protective
measures are necessary.
(ii) If the application is for an OL or
COL or if the application is for an early
site permit and contains plans for
coping with emergencies under
§ 53.1146(b)(2)(ii), and if the plume
exposure pathway EPZ extends beyond
the site boundary:
(A) The applicant must submit
radiological emergency response plans
of State, local, and participating Tribal
governmental entities in the United
States that are wholly or partially within
the plume exposure pathway EPZ.
(B) The exact configuration of the
plume exposure pathway EPZ
surrounding the facility shall be
determined in relation to the local
emergency response needs and
capabilities as they are affected by such
conditions as demography, topography,
land characteristics, access routes, and
jurisdictional boundaries.
(iii) If the application is for an early
site permit that, under § 53.1146(b)(2)(i),
proposes major features of the
emergency plans and describes the EPZ,
and if the EPZ extends beyond the site
boundary, then the exact configuration
of the plume exposure pathway EPZ
surrounding the facility must be
determined in relation to the local
emergency response needs and
capabilities as they are affected by such
conditions as demography, topography,
land characteristics, access routes, and
jurisdictional boundaries.
(h) [Reserved]
(i) A list of the names and addresses
of such regulatory agencies as may have
jurisdiction over the rates and services

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incident to the proposed activity, and a
list of trade and news publications
which circulate in the area where the
proposed activity will be conducted and
which are considered appropriate to
give reasonable notice of the application
to those municipalities, private utilities,
public bodies, and cooperatives, which
might have a potential interest in the
facility; and
(j) If the application contains
Restricted Data or classified National
Security information, confirmation that
all Restricted Data and classified
National Security information are
separated from the unclassified
information.
1 EPZs are discussed in NUREG–0396, U.S.
Environmental Protection Agency 520/1–78–
016, ‘‘Planning Basis for the Development of
State and Local Government Radiological
Emergency Response Plans in Support of
Light-Water Nuclear Power Plants,’’
December 1978.
2 If the State, local, and participating Tribal
emergency response plans have been
previously provided to the NRC for inclusion
in the facility docket, the applicant need only
provide the appropriate reference to meet
this requirement.

§ 53.1112

Environmental conditions.

(a) Each CP, early site permit, and
COL under this part may include
conditions to address environmental
issues during construction. These
conditions are to be set out in an
attachment to the license, which is
incorporated in and made a part of the
license. These conditions will be
derived from information contained in
the environmental report submitted
pursuant to § 51.50 of this chapter, as
analyzed and evaluated in the NRC
record of decision and will identify the
obligations of the licensee in the
environmental area, including, as
appropriate, requirements for reporting
and keeping records of environmental
data, and any conditions and
monitoring requirement for the
protection of the nonaquatic
environment.
(b) Each license authorizing operation
of a commercial nuclear plant under
this part, and each license for a
commercial nuclear plant for which the
certification of permanent cessation of
operations required under § 53.1070 has
been submitted may include conditions
to address environmental issues during
operation and decommissioning. These
conditions are to be set out in an
attachment to the license, which is
incorporated in and made a part of the
license. These conditions will be
derived from information contained in
the environmental report or the
supplement to the environmental report
submitted under §§ 51.50 and 51.53 of

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this chapter as analyzed and evaluated
in the NRC record of decision, and will
identify the obligations of the licensee
in the environmental area, including, as
appropriate, requirements for reporting
and keeping records of environmental
data and any conditions and monitoring
requirement for the protection of the
nonaquatic environment.
§ 53.1115 Agreement limiting access to
classified information.

As part of its application and in any
event before the receipt of Restricted
Data or classified National Security
Information or the issuance of a license
or standard design approval under this
part, or before the Commission has
adopted a final standard design
certification rule under this part, the
applicant must agree in writing that it
will not permit any individual to have
access to or any facility to possess
Restricted Data or classified National
Security Information until the
individual and/or facility has been
approved for access under the
provisions of 10 CFR parts 25 and/or 95.
The agreement of the applicant becomes
part of the license or standard design
approval.
§ 53.1118

Ineligibility of certain applicants.

Any person who is a citizen, national,
or agent of a foreign country, or any
corporation, or other entity which the
Commission knows or has reason to
believe is owned, controlled, or
dominated by an alien, a foreign
corporation, or a foreign government,
will be ineligible to apply for and obtain
a license.
§ 53.1120 Exceptions and exemptions
from licensing requirements.

Nothing in this part must be deemed
to require a license for—
(a) The manufacture, production, or
acquisition by the Department of
Defense of any utilization facility
authorized pursuant to section 91 of the
Act or the use of such facility by the
Department of Defense or by a person
under contract with and for the account
of the Department of Defense;
(b) Except to the extent that the
Department of Energy facilities of the
types subject to licensing pursuant to
section 202 of the Energy
Reorganization Act of 1974, as
amended, are involved—
(1)(i) The processing, fabrication or
refining of special nuclear material
(SNM) or the separation of SNM, or the
separation of SNM from other
substances by a prime contractor of the
Department of Energy under a prime
contract for—

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(A) The performance of work for the
Department of Energy at a United States
government-owned or controlled site;
(B) Research in, or development,
manufacture, storage, testing or
transportation of, atomic weapons or
components thereof; or
(C) The use or operation of a
utilization facility in a United States
owned vehicle or vessel; or
(ii) The processing, fabrication or
refining of SNM of the separation of
SNM, or the separation of SNM from
other substances by a prime contractor
or subcontractor of the Commission or
the Department of Energy under a prime
contract or subcontract when the
Commission determines that the
exemption of the prime contractor or
subcontractor is authorized by law; and
that, under the terms of the contract or
subcontract, there is adequate assurance
that the work thereunder can be
accomplished without undue risk to the
public health and safety; or
(2)(i) The construction or operation of
a utilization facility for the Department
of Energy at a United States
government-owned or controlled site,
including the transportation of the
utilization facility to or from such site
and the performance of contract services
during temporary interruptions of such
transportation; or the construction or
operation of a utilization facility for the
Department of Energy in the
performance of research in, or
development, manufacture, storage,
testing, or transportation of, atomic
weapons or components thereof; or the
use or operation of a utilization facility
for the Department of Energy in a
United States government-owned
vehicle or vessel; provided that such
activities are conducted by a prime
contractor of the Department of Energy
under a prime contract with the
Department of Energy; or
(ii) The construction or operation of a
utilization facility by a prime contractor
or subcontractor of the Commission or
the Department of Energy under his or
her prime contract or subcontract when
the Commission determines that the
exemption of the prime contractor or
subcontractor is authorized by law; and
that, under the terms of the contract or
subcontract, there is adequate assurance
that the work thereunder can be
accomplished without undue risk to the
public health and safety; or
(c) The transportation or possession of
any utilization facility by a common or
contract carrier or warehouse employee
in the regular course of carriage for
another or storage incident thereto.

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§ 53.1121 Public inspection of
applications.

Applications and documents
submitted to the Commission in
connection with applications may be
made available for public inspection
under the provisions of part 2 of this
chapter.
§ 53.1124

Relationship between sections.

(a) Limited work authorization. An
application for a limited work
authorization (LWA) under this part
may be submitted as part of an
application for an early site permit, CP,
or COL under this part as required in
§ 53.1130(a)(2).
(b) Early site permit. (1) A holder of
an early site permit may request an
LWA.
(2) An application for a CP or COL
under this part may, but need not,
reference an early site permit.
(c) Standard design approval. An
application for a standard design
approval under this part may, but need
not, reference an OL or custom COL
under this part that is essentially the
same as the information supporting the
standard design for which approval is
being requested.
(d) Standard design certification. An
application for a standard design
certification under this part may, but
need not, reference an OL or custom
COL under this part that is essentially
the same as the standard design for
which certification is being requested.
(e) Manufacturing license. (1) A
manufactured reactor manufactured
under an ML issued under this part may
only be transported to and installed at
a site for which a COL under this part
has been issued.
(2) An ML applicant under this part
may reference a standard design
certification or a standard design
approval under this part in its
application.
(f) Construction permit. An
application for a CP may, but need not,
reference a standard design certification
or standard design approval issued
under this part, respectively, and may
also reference an early site permit
issued under this part. In the absence of
a demonstration that an entity other
than the one originally sponsoring a
standard design certification is qualified
to supply a design, the Commission will
entertain an application for a CP that
references a standard design
certification issued under this part only
if the entity that sponsored the
certification supplies the design for the
applicant’s use.
(g) Operating license. (1) An
application for an OL under this part
may, but need not, reference an early

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site permit, standard design
certification, or standard design
approval issued under this part. In the
absence of a demonstration that an
entity other than the one originally
sponsoring a standard design
certification is qualified to supply a
design, the Commission will entertain
an application for an OL that references
a standard design certification issued
under this part only if the entity that
sponsored the certification supplies the
design for the applicant’s use.
(2) The holder of a CP must, at the
time of submission of the Final Safety
Analysis Report (FSAR), file an
application for an OL.
(h) Combined licenses. An application
for a COL under this part may, but need
not, reference an early site permit,
standard design certification, standard
design approval, or ML issued under
this part. In the absence of a
demonstration that an entity other than
the one originally sponsoring and
obtaining a standard design certification
is qualified to supply a design, the
Commission will entertain an
application for a COL that references a
standard design certification issued
under this part only if the entity that
sponsored the certification supplies the
design for the applicant’s use.
§ 53.1130

Limited work authorizations.

(a) Request for limited work
authorization. (1) Any person to whom
the Commission may otherwise issue
either a license or permit related to a
commercial nuclear plant may request
an LWA allowing that person to perform
the driving of piles, subsurface
preparation, placement of backfill,
concrete, or permanent retaining walls
within an excavation, and installation of
the foundation, including placement of
concrete, any of which are for a
structure, system, or component (SSC)
of the facility for which either a CP or
COL is otherwise required under
§ 53.610.
(2) An application for an LWA may be
submitted as part of a complete
application for a CP or COL in
accordance with § 2.101(a)(1) through
(a)(5) of this chapter, or as a partial
application in accordance with
§ 2.101(a)(9) of this chapter. An
application for an LWA by the holder of
an early site permit must be submitted
as a complete application in accordance
with § 2.101(a)(1) through (a)(4) of this
chapter.
(3) The application must include—
(i) A Safety Analysis Report required
by §§ 53.1146, 53.1309 or 53.1416, as
applicable, a description of the activities
requested to be performed, and the
design and construction information

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otherwise required by the Commission’s
rules and regulations to be submitted for
a CP or COL under this part but limited
to those portions of the facility that are
within the scope of the LWA. The Safety
Analysis Report must demonstrate that
activities conducted under the LWA
will be conducted in compliance with
the technically relevant Commission
requirements in 10 CFR chapter I
applicable to the design of those
portions of the facility within the scope
of the LWA;
(ii) An environmental report in
accordance with § 51.49 of this chapter;
and
(iii) A plan for redress of activities
performed under the LWA, should
limited work activities be terminated by
the holder, or the LWA be revoked by
the NRC or upon effectiveness of the
Commission’s final decision denying
the associated CP or COL application, as
applicable.
(b) Issuance of limited work
authorization. (1) The Director, Office of
Nuclear Reactor Regulation may issue
an LWA only after—
(i) The NRC staff issues the final
environmental impact statement for the
LWA under subpart A of part 51 of this
chapter;
(ii) The presiding officer makes the
finding in §§ 51.105(c) or 51.107(d) of
this chapter, as applicable;
(iii) The Director determines that the
applicable standards and requirements
of the Act, and the Commission’s
regulations applicable to the activities to
be conducted under the LWA, have
been met, the applicant is technically
qualified to engage in the activities
authorized, and that issuance of the
LWA will provide reasonable assurance
of adequate protection to public health
and safety and will not be inimical to
the common defense and security; and
(iv) The presiding officer finds that
there are no unresolved safety issues
relating to the activities to be conducted
under the LWA that would constitute
good cause for withholding the
authorization.
(2) Each LWA will specify the
activities that the holder is authorized to
perform.
(c) Effect of limited work
authorization. Any activities
undertaken under an LWA are entirely
at the risk of the applicant and, except
as to the matters determined under
paragraph (b)(1) of this section, the
issuance of the LWA has no bearing on
the issuance of a CP or COL with respect
to the requirements of the Act and rules,
regulations, or orders issued under the
Act. The environmental impact
statement for a CP or COL application
for which an LWA was previously

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issued will not address, and the
presiding officer will not consider, the
sunk costs of the holder of the LWA in
determining the proposed action (i.e.,
issuance of the CP or COL).
(d) Implementation of redress plan. If
construction is terminated by the
holder, the underlying application is
withdrawn by the applicant or denied
by the NRC, or the LWA is revoked by
the NRC, then the holder must begin
implementation of the redress plan in a
reasonable time. The holder must also
complete the redress of the site no later
than 18 months after termination of
construction, revocation of the LWA, or
upon effectiveness of the Commission’s
final decision denying the associated CP
application or the associated COL
application, as applicable.
§ 53.1140

Early site permits.

Sections 53.1140 through 53.1188 set
out the requirements and procedures
applicable to Commission issuance of
an early site permit under this part for
approval of a site for a commercial
nuclear plant separate from the filing of
an application for a CP or COL for the
facility.
§ 53.1143

Filing of applications.

Any person who may apply for a CP
or for a COL under this part, may file
an application for an early site permit
with the Director, Office of Nuclear
Reactor Regulation. An application for
an early site permit may be filed
notwithstanding the fact that an
application for a CP or a COL has not
been filed in connection with the site
for which a permit is sought.
§ 53.1144 Contents of applications for
early site permits; general information.

The application must contain all of
the information required by § 53.1109(a)
through (d) and (j).
§ 53.1146 Contents of applications for
early site permits; technical information.

(a) The application must contain—
(1) A Site Safety Analysis Report that
must include the following:
(i) The specific number, type, and
thermal power level of the facilities, or
range of possible facilities, for which the
site may be used;
(ii) The anticipated maximum levels
of radiological and thermal effluents
each facility will produce;
(iii) The type of cooling systems,
including intakes and outflows, where
appropriate, that may be associated with
each facility;
(iv) The boundaries of the site;
(v) The proposed general location of
each facility on the site;
(vi) The external hazards and site
characteristics required by this part;

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(vii) The location and description of
any nearby industrial, military, or
transportation facilities and routes;
(viii) The existing and projected
future population profile of the area
surrounding the site;
(ix) A description and assessment of
the site on which a facility is to be
located. The assessment must address
the requirements of subpart D of this
part;
(x) Information demonstrating that
site characteristics are such that
adequate security plans and measures
can be developed; and
(xi) A description of the quality
assurance program (QAP) required by
appendix B to part 50 of this chapter
applied to site-related activities for the
future design, fabrication, construction,
and testing of the SSCs of a facility or
facilities that may be constructed on the
site.
(2) A complete environmental report
as required by § 51.50(b) of this chapter.
(b)(1) The Site Safety Analysis Report
must identify physical characteristics of
the proposed site, such as egress
limitations from the area surrounding
the site, that could pose a significant
impediment to the development of
emergency plans. If physical
characteristics are identified that could
pose a significant impediment to the
development of emergency plans, the
application must identify measures that
would, when implemented, mitigate or
eliminate the significant impediment.
(2) The Site Safety Analysis Report
may also—
(i) Propose major features of the
emergency plans, under either § 50.160
or the requirements in appendix E to
part 50 and § 50.47(b) of this chapter, as
applicable, such as the exact size and
configuration of the EPZs, for review
and approval by the NRC, in
consultation with the Federal
Emergency Management Agency
(FEMA), as applicable, in the absence of
complete and integrated emergency
plans; or
(ii) Propose complete and integrated
emergency plans for review and
approval by the NRC, in consultation
with FEMA, as applicable, in
accordance with either § 50.160 or the
requirements in appendix E to part 50
and § 50.47(b) of this chapter. To the
extent approval of emergency plans is
sought, the application must contain the
information required by § 53.1109(g).
(3) Emergency plans submitted under
paragraph (b)(2)(ii) of this section must
include the proposed inspections, tests,
and analyses that the holder of a COL
referencing the early site permit must
perform, and the acceptance criteria that
are necessary and sufficient to provide

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reasonable assurance that, if the
inspections, tests, and analyses are
performed and the acceptance criteria
met, the facility has been constructed
and will be operated in conformity with
the emergency plans, the provisions of
the Act, and the Commission’s rules and
regulations. Major features of an
emergency plan submitted under
paragraph (b)(2)(i) of this section may
include proposed inspections, tests,
analyses, and acceptance criteria
(ITAAC).
(4) Under paragraphs (b)(1) and
(b)(2)(i) of this section, the Site Safety
Analysis Report must include, where
appropriate, a description of contacts
and arrangements made with Federal,
State, participating Tribal, and local
governmental agencies with emergency
planning responsibilities. The Site
Safety Analysis Report must contain any
certifications that have been obtained. If
these certifications, where appropriate,
cannot be obtained, the Site Safety
Analysis Report must contain
information, including a utility plan,
sufficient to show that the proposed
plans provide reasonable assurance that
adequate protective measures can and
will be taken in the event of a
radiological emergency at the site.
Under the option set forth in paragraph
(b)(2)(ii) of this section, the applicant
must make good faith efforts, where
appropriate, to obtain from the same
governmental agencies certifications
that—
(i) The proposed emergency plans are
practicable;
(ii) These agencies are committed to
participating in any further
development of the plans, including any
required field demonstrations; and
(iii) That these agencies are
committed to executing their
responsibilities under the plans in the
event of an emergency.
(c) An applicant may request that an
LWA under § 53.1130 be issued in
conjunction with the early site permit.
The application must include the
information otherwise required by
§ 53.1130.
(d) Each applicant for an early site
permit under this part must protect
safeguards information against
unauthorized disclosure in accordance
with the requirements in §§ 73.21 and
73.22 of this chapter, as applicable.
§ 53.1149

Review of applications.

(a) Standards for review of
applications. Applications filed under
this part will be reviewed according to
the applicable standards set out in this
part. In addition, the Commission must
prepare an environmental impact
statement during review of the

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application, under the applicable
provisions of 10 CFR part 51. The
Commission must determine, after
consultation with FEMA, as applicable,
whether the information required of the
applicant by § 53.1146(b)(1) shows that
there is no significant impediment to
the development of emergency plans
that cannot be mitigated or eliminated
by measures proposed by the applicant,
whether any major features of
emergency plans submitted by the
applicant under § 53.1146(b)(2)(i) are
acceptable under either § 50.160 or
appendix E to part 50 and § 50.47(b) of
this chapter, and whether any
emergency plans submitted by the
applicant under § 53.1146(b)(2)(ii)
provide reasonable assurance that
adequate protective measures can and
will be taken in the event of a
radiological emergency.
(b) Administrative review of
applications; hearings. An early site
permit application is subject to all
procedural requirements in 10 CFR part
2, including the requirements for
docketing in § 2.101(a)(1) through (4) of
this chapter, and the requirements for
issuance of a notice of hearing in
§ 2.104(a) and (d) of this chapter,
provided that the designated sections
may not be construed to require that the
environmental report, or draft or final
environmental impact statement
includes an assessment of the benefits of
construction and operation of the
reactor or reactors, or an analysis of
alternative energy sources. The
presiding officer in an early site permit
hearing must not admit contentions
proffered by any party concerning an
assessment of the benefits of
construction and operation of the
reactor or reactors, or an analysis of
alternative energy sources if those issues
were not addressed by the applicant in
the early site permit application. All
hearings conducted on applications for
early site permits filed under this part
are governed by the procedures
contained in subparts C, G, L, and N of
10 CFR part 2, as applicable.

Commission deems appropriate, if the
Commission finds that—
(1) An application for an early site
permit demonstrates compliance with
the applicable standards and
requirements of the Act and the
Commission’s regulations;
(2) Notifications, if any, to other
agencies or bodies have been duly
made;
(3) There is reasonable assurance that
the site is in conformity with the
provisions of the Act and the
Commission’s regulations;
(4) The applicant is technically
qualified to engage in any activities
authorized;
(5) The proposed ITAAC, including
any on emergency planning, are
necessary and sufficient, within the
scope of the early site permit, to provide
reasonable assurance that the facility
has been constructed and will be
operated in conformity with the license,
the provisions of the Act, and the
Commission’s regulations;
(6) Issuance of the permit will not be
inimical to the common defense and
security or to the health and safety of
the public;
(7) Any significant adverse
environmental impact resulting from
activities requested under § 53.1146(c)
can be redressed; and
(8) The findings required by subpart
A of 10 CFR part 51 have been made.
(b) The early site permit must specify
the site characteristics, design
parameters, and terms and conditions of
the early site permit the Commission
deems appropriate. Before issuance of
either a CP or COL referencing an early
site permit, the Commission must find
that any relevant terms and conditions
of the early site permit have been met.
Any terms or conditions of the early site
permit that could not be met by the time
of issuance of the CP or COL, must be
set forth as terms or conditions of the CP
or COL.
(c) The early site permit must specify
those § 53.1130(b) activities requested
under § 53.1146(c) that the permit
holder is authorized to perform.

§ 53.1155 Referral to the Advisory
Committee on Reactor Safeguards.

§ 53.1161

The Commission must refer a copy of
the application for an early site permit
to the Advisory Committee on Reactor
Safeguards (ACRS). The ACRS must
report on those portions of the
application which concern safety.
§ 53.1158

Issuance of early site permit.

(a) After conducting a hearing under
§ 53.1149(b) and receiving the report to
be submitted by the ACRS under
§ 53.1155, the Commission may issue an
early site permit, in the form the

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Extent of activities permitted.

If the activities authorized by
§ 53.1158(c) are performed and the site
is not referenced in an application for a
CP or a COL issued under this part
while the permit remains valid, then the
early site permit remains in effect solely
for the purpose of site redress, and the
holder of the permit must redress the
site under the terms of the site redress
plan required by § 53.1146(c). If, before
redress is complete, a use not envisaged
in the redress plan is found for the site
or parts thereof, the holder of the permit

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must carry out the redress plan to the
greatest extent possible consistent with
the alternate use.
§ 53.1164

Duration of permit.

(a) Except as provided in paragraph
(b) of this section, an early site permit
issued under this subpart may be valid
for not less than 10, nor more than 20
years from the date of issuance.
(b) An early site permit continues to
be valid beyond the date of expiration
in any proceeding on a CP application
or a COL application that references the
early site permit and is docketed before
the date of expiration of the early site
permit, or, if a timely application for
renewal of the permit has been
docketed, before the Commission has
determined whether to renew the
permit.
(c) An applicant for a CP or COL may,
at its own risk, reference in its
application a site for which an early site
permit application has been docketed
but not granted.
(d) Upon issuance of a CP or COL, a
referenced early site permit is
subsumed, to the extent referenced, into
the CP or COL.
§ 53.1167 Limited work authorization after
issuance of early site permit.

A holder of an early site permit may
request an LWA under § 53.1130.
§ 53.1170

Transfer of early site permit.

An application to transfer an early site
permit will be processed under
§ 53.1570.

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§ 53.1173

Application for renewal.

(a) Not less than 12, nor more than 36
months before the expiration date stated
in the early site permit, or any later
renewal period, the permit holder may
apply for a renewal of the permit. An
application for renewal must contain all
information necessary to bring up to
date the information and data contained
in the previous application.
(b) Any person whose interests may
be affected by renewal of the permit
may request a hearing on the
application for renewal. The request for
a hearing must comply with § 2.309 of
this chapter. If a hearing is granted,
notice of the hearing will be published
under § 2.309 of this chapter.
(c) An early site permit, either original
or renewed, for which a timely
application for renewal has been filed,
remains in effect until the Commission
has determined whether to renew the
permit. If the permit is not renewed, it
continues to be valid in certain
proceedings in accordance with the
provisions of § 53.1164(b).
(d) The Commission must refer a copy
of the application for renewal to the

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ACRS. The ACRS must report on those
portions of the application which
concern safety and must apply the
criteria set forth in § 53.1176.
§ 53.1176

Criteria for renewal.

(a) The Commission must grant the
renewal if it determines that—
(1) The site complies with the Act, the
Commission’s regulations, and orders
applicable and in effect at the time the
site permit was originally issued; and
(2) Any new requirements the
Commission may wish to impose—
(i) Are necessary for adequate
protection to public health and safety or
common defense and security;
(ii) Are necessary for compliance with
the Commission’s regulations, and
orders applicable and in effect at the
time the site permit was originally
issued; or
(iii) Would provide a substantial
increase in overall protection of the
public health and safety or the common
defense and security to be derived from
the new requirements, and the direct
and indirect costs of implementation of
those requirements are justified in view
of this increased protection.
(b) A denial of renewal under the
provisions of § 53.1176(a) does not bar
the permit holder or another applicant
from filing a new application for the site
which proposes changes to the site or
the way that it is used to correct the
deficiencies cited in the denial of the
renewal.
§ 53.1179

Duration of renewal.

Each renewal of an early site permit
may be for not less than 10, nor more
than 20 years, plus any remaining years
on the early site permit then in effect
before renewal.
§ 53.1182

Use of site for other purposes.

A site for which an early site permit
has been issued under this part may be
used for purposes other than those
described in the permit, including the
location of other types of energy
facilities. The permit holder must
inform the Director, Office of Nuclear
Reactor Regulation (Director), of any
significant uses for the site which have
not been approved in the early site
permit. The information about the
activities must be given to the Director
at least 30 days in advance of any actual
construction or site modification for the
activities. The information provided
could be the basis for imposing new
requirements on the permit, under the
provisions of § 53.1188. If the permit
holder informs the Director that the
holder no longer intends to use the site
for a commercial nuclear plant, the
Director may terminate the permit.

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§ 53.1188 Finality of early site permit
determinations.

(a) Commission finality. (1) While an
early site permit is in effect under
§ 53.1164 or § 53.1179, the Commission
may not change or impose new site
characteristics, design parameters, or
terms and conditions, including
emergency planning requirements, on
the early site permit unless the
Commission—
(i) Determines that a modification is
necessary to bring the permit or the site
into compliance with the Commission’s
regulations and orders applicable and in
effect at the time the permit was issued;
(ii) Determines the modification is
necessary to assure adequate protection
of the public health and safety or the
common defense and security;
(iii) Determines that a modification is
necessary based on an update under
paragraph (b) of this section; or
(iv) Issues a variance requested under
paragraph (d) of this section.
(2) In making the findings required for
issuance of a CP, COL, or OL, or the
findings required by § 53.1452(g), or in
any enforcement hearing other than one
initiated by the Commission under
paragraph (a)(1) of this section, if the
application for the CP, COL, or OL
references an early site permit, the
Commission must treat as resolved
those matters resolved in the proceeding
on the application for issuance or
renewal of the early site permit, except
as provided for in paragraphs (b), (c),
and (d) of this section.
(i) If the Commission grants a CP
application that references an early site
permit and an application for an OL
references the CP, the Commission must
treat as resolved those matters resolved
in the proceeding for the issuance or
renewal of the early site permit, except
as provided for in paragraphs (b), (c),
and (d) of this section.
(ii) If the early site permit approved
an emergency plan (or major features
thereof) that is in use by a licensee of
a commercial nuclear plant, the
Commission must treat as resolved
changes to the early site permit
emergency plan (or major features
thereof) that are identical to changes
made to the licensee’s emergency plans
under § 53.1565 occurring after issuance
of the early site permit.
(iii) If the early site permit approved
an emergency plan (or major features
thereof) that is not in use by a licensee
of a commercial nuclear plant, the
Commission must treat as resolved
changes that are equivalent to those that
could be made under § 53.1565 without
prior NRC approval had the emergency
plan been in use by a licensee.

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(b) Updating of early site permitemergency preparedness. An applicant
for a CP, OL, or COL who has filed an
application referencing an early site
permit issued under this subpart must
update the emergency preparedness
information that was provided under
§ 53.1146(b) and discuss whether the
updated information materially changes
the bases for compliance with
applicable NRC requirements.
(c) Hearings and petitions. (1) In any
proceeding for the issuance of a CP, OL,
or COL referencing an early site permit,
contentions on the following matters
may be litigated in the same manner as
other issues material to the proceeding:
(i) The nuclear reactor proposed to be
built does not fit within one or more of
the site characteristics or design
parameters included in the early site
permit;
(ii) One or more of the terms and
conditions of the early site permit have
not been met;
(iii) A variance requested under
paragraph (d) of this section is
unwarranted or should be modified;
(iv) New or additional information is
provided in the application that
substantially alters the bases for a
previous NRC conclusion or constitutes
a sufficient basis for the Commission to
modify or impose new terms and
conditions related to emergency
preparedness; or
(v) Any significant environmental
issue that was not resolved in the early
site permit proceeding, or any issue
involving the impacts of construction
and operation of the facility that was
resolved in the early site permit
proceeding for which significant new
information has been identified.
(2) Any person may file a petition
requesting that the site characteristics,
design parameters, or terms and
conditions of the early site permit be
modified, or that the permit be
suspended or revoked. The petition will
be considered under § 2.206 of this
chapter. Before construction
commences, the Commission must
consider the petition and determine
whether any immediate action is
required. If the petition is granted, an
appropriate order will be issued.
Construction under the CP or COL will
not be affected by the granting of the
petition unless the order is made
immediately effective. Any change
required by the Commission in response
to the petition must demonstrate
compliance with the requirements of
paragraph (a)(1) of this section.
(d) Variances. An applicant for a CP,
OL, or COL referencing an early site
permit may include in its application a
request for a variance from one or more

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site characteristics, design parameters,
or terms and conditions of the early site
permit, or from the Site Safety Analysis
Report. In determining whether to grant
the variance, the Commission must
apply the same technically relevant
criteria applicable to the application for
the original or renewed early site
permit. Once a CP or COL referencing
an early site permit is issued, variances
from the early site permit will not be
granted for that CP or COL.
(e) Early site permit amendment. The
holder of an early site permit may not
make changes to the early site permit or
the Site Safety Analysis Report without
prior Commission approval. The request
for a change to the early site permit
must be in the form of an application for
a license amendment and must
demonstrate compliance with the
requirements of §§ 53.1510 and 53.1520.
§ 53.1200

Standard design approvals.

Sections 53.1200 through 53.1221 set
out procedures for the filing, NRC staff
review, and referral to the ACRS of
standard designs, or major portions
thereof, for a commercial nuclear plant
under this part.
§ 53.1203

Filing of applications.

Any person may submit a proposed
standard design for a commercial
nuclear plant to the NRC staff for its
review. The submittal may consist of
either the final design for the entire
facility or the final design for major
portions thereof.
§ 53.1206 Contents of applications for
standard design approvals; general
information.

The application must contain all of
the information required by § 53.1109(a)
through (c) and (j).
§ 53.1209 Contents of applications for
standard design approvals; technical
information.

(a) Major portion of a standard design.
If the applicant seeks review of a major
portion of a standard design, the
application need only contain the
information required by this section to
the extent the requirements are
applicable to the major portion of the
standard design for which NRC staff
approval is sought. If an applicant seeks
approval of a major portion of the
design, the scope of the application for
which approval is sought must include
all functional design criteria necessary
to demonstrate compliance with the
safety criteria in §§ 53.210, 53.220 and
53.450(e), as applicable, for the major
portion of the standard design for which
NRC staff approval is sought. Such
applicants must identify conditions
related to interfaces with systems

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outside the scope of the major portion
of the standard design for which NRC
staff approval is sought, and functional
or physical boundary conditions
between the major portion of the
standard design for which NRC staff
approval is sought and the remainder of
the standard design. These conditions
must be demonstrated when the
standard design approval is
incorporated into a subsequent CP,
design certification, ML, or COL
application.
(b) Final Safety Analysis Report. The
application must contain an FSAR that
describes the facility and the limits on
its operation, presents a safety analysis
of the SSCs and of the facility, or major
portions thereof, for which the applicant
seeks design approval, and must include
the following information:
(1) Site Parameters. The site
parameters postulated for the design
under this part, including the designbasis external hazard levels for the
relevant external hazards, and an
analysis and evaluation of the design in
terms of those site parameters.
(2) Design information. Except as
specified in this paragraph, an
application for a standard design
approval for a commercial nuclear plant
must include the design information
equivalent to that required for a
standard design certification under
§ 53.1239(a)(2) through (27) for those
portions of a commercial nuclear plant
included in the standard design
approval.
§ 53.1210 Contents of applications for
standard design approvals; other
application content.

(a) In addition to the FSAR, the
application must also include the
following:
(1) Availability Controls (if not
included in the FSAR). A description of
the controls on plant operations,
including availability controls, to
provide reasonable confidence that the
configurations and special treatments
for safety-related (SR) SSCs and nonsafety-related but safety-significant
(NSRSS) SSCs provide the capabilities
and reliabilities required to demonstrate
compliance with the safety criteria of
§ 53.220.
(2) Safeguards Information. A
description of the program to protect
Safeguards Information against
unauthorized disclosure in accordance
with the requirements in §§ 73.21 and
73.22 of this chapter, as applicable.
(b) If there are SSCs of the plant
which required research and
development to confirm the adequacy of
their design, provide a report in the
application which documents the

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resolution of any safety questions
associated with such SSCs.
(c) A description of how the
performance of each design feature has
been demonstrated capable of fulfilling
functional design criteria considering
interdependent effects through either
analysis, appropriate test programs,
prototype testing, operating experience,
or a combination thereof, in accordance
with § 53.440(a).
§ 53.1212 Standards for review of
applications.

Applications filed under this part will
be reviewed under the standards set out
in 10 CFR parts 20, 53, and 73.
§ 53.1215 Referral to the Advisory
Committee on Reactor Safeguards.

The Commission must refer a copy of
the application to the ACRS. The ACRS
must report on those portions of the
application which concern safety.
§ 53.1218

Staff approval of design.

(a) Upon completion of its review of
a submittal under §§ 53.1200 through
53.1221 and receipt of a report by the
ACRS under § 53.1215, the NRC staff
must publish a determination in the
Federal Register as to whether or not
the design is acceptable, subject to
appropriate terms and conditions, and
make an analysis of the design in the
form of a report available at the NRC
website, https://www.nrc.gov.
(b) A standard design approval issued
under this section is valid for 15 years
from the date of issuance and may not
be renewed. A design approval
continues to be valid beyond the date of
expiration in any proceeding on an
application for a CP, OL, COL, or ML
under this part that references the
design approval and is docketed before
the date of expiration of the design
approval.

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§ 53.1221 Finality of standard design
approvals; information requests.

(a) An approved design must be used
by and relied upon by the NRC staff and
the ACRS in their reviews of any
standard design certification or
individual facility license application
under this part that incorporates by
reference a standard design approved
under this part unless there exists
significant new information that
substantially affects the earlier
determination or other good cause.
(b) The determination and report by
the NRC staff do not constitute a
commitment to issue a permit or
license, or in any way affect the
authority of the Commission, Atomic
Safety and Licensing Board Panel, or
presiding officers in any proceeding
under part 2 of this chapter.

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(c) Except for information requests
seeking to verify compliance with the
current licensing basis of the standard
design approval, information requests to
the holder of a standard design approval
must be evaluated before issuance to
ensure that the burden to be imposed on
respondents is justified in view of the
potential safety significance of the issue
to be addressed in the requested
information. Each evaluation performed
by the NRC staff must be in accordance
with § 53.1580 and must be approved by
the Executive Director for Operations or
authorized designee before issuance of
the request.
(d) The Commission will require,
before granting a CP, COL, OL, or ML
that references a standard design
approval, that engineering documents,
such as analyses, drawings,
procurement specifications, or
construction and installation
specifications, be completed and
available for audit if the more detailed
information is necessary for the
Commission to verify the information in
the application and make its safety
determination, including the
determination that the application is
consistent with the design approval
information. This information may be
acquired by appropriate arrangements
with the design approval applicant.
§ 53.1230

Standard design certifications.

Sections 53.1230 through 53.1263 set
forth the requirements and procedures
applicable to the Commission’s issuance
of rules granting standard design
certifications for commercial nuclear
plants under this part separate from the
filing of an application for a CP or COL
for such a facility.
§ 53.1233

Filing of applications.

(a) An application for design
certification may be filed
notwithstanding the fact that an
application for a CP, COL, or ML for
such a facility has not been filed.
(b) The application must comply with
the applicable filing requirements of
§ 53.040 and §§ 2.811 through 2.819 of
this chapter.
§ 53.1236 Contents of applications for
standard design certifications; general
information.

The application must contain all of
the information required by § 53.1109(a)
through (c) and (j).
§ 53.1239 Contents of applications for
standard design certifications; technical
information.

The application must contain a level
of design information sufficient to
enable the Commission to judge the
applicant’s proposed means of assuring

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87085

that construction conforms to the design
and to reach a final conclusion on all
safety questions associated with the
design before the certification is
granted. The information submitted for
a design certification must include
performance requirements and design
information sufficiently detailed to
permit the preparation of acceptance
and inspection requirements by the
NRC. The Commission will require,
before design certification, that
information normally contained in
engineering documents, such as
analyses, drawings, procurement
specifications, or construction and
installation specifications, be completed
and available for audit if the more
detailed information is necessary for the
Commission to verify the information in
the application and make its safety
determination.
(a) Final Safety Analysis Report. The
application must contain an FSAR that
describes the facility and the limits on
its operation, and presents a safety
analysis of the SSCs, and must include
the following information:
(1) Site Parameters. The site
parameters postulated for the design
under this part, including the designbasis external hazard levels for the
relevant external hazards, and an
analysis and evaluation of the design in
terms of those site parameters.
(2) Plant Description and Safety
Functions—(i) General Plant
Description. A general description of the
commercial nuclear plant including
reactor type, the intended use of the
reactor, nuclear design (e.g., neutron
spectrum, reactor control, multi-unit
reactor control), overall layout of the
plant including significant plant
features and SSCs, maximum power
level and the nature and inventory of
radioactive materials.
(ii) Safety functions. A description of
the primary and additional safety
functions required under § 53.230 and a
summary of how each safety function is
satisfied.
(3) Design Features and functional
design criteria—licensing-basis events.
(i) A description of the design features
required by § 53.400 and the functional
design criteria required by §§ 53.410
and 53.420 that, when combined with
corresponding human actions and
programmatic controls, demonstrate that
the plant will demonstrate compliance
with the safety criteria defined in
§ 53.210 and established in accordance
with § 53.220, or more restrictive
alternative criteria adopted under
§ 53.470, during licensing-basis events
(LBEs).
(ii) A description of how design
features demonstrate compliance with

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the requirements of § 53.440(a) through
(i) and (k) through (m).
(4) Design Features Supporting
Normal Operations. A description of the
design features required by § 53.425 to
support the holder of an OL or COL
complying with § 53.260 during normal
operations.
(5) Design Features and Functional
Design Criteria—aircraft impact. A
description of the design features and
functional design criteria required to
demonstrate compliance with the
requirements of § 53.440(j) for
addressing the impact of a large,
commercial aircraft.
(6) Earthquake engineering. The
information necessary to demonstrate
that the commercial nuclear plant
complies with the earthquake
engineering criteria in § 53.480.
(7) Programmatic Controls and
Interfaces. (i) A description of the
corresponding programmatic controls
and interfaces necessary to achieve and
maintain the reliability and capability of
SSCs relied upon to demonstrate
compliance with the functional design
criteria required by §§ 53.410 and
53.420 and the safety criteria in
§§ 53.210 and 53.220, or more restrictive
alternative criteria adopted under
§ 53.470, and necessary to maintain
consistency with analyses required by
§ 53.450.
(ii) For an application for a multi-unit
commercial nuclear plant, the
programmatic controls and interfaces
must also be described for different
modular configurations, as required by
§ 53.440(i), including any restrictions
that will be necessary during the
construction and startup of any given
unit to ensure the safe operation of the
overall commercial nuclear plant to be
licensed under this part.
(8) Programmatic Controls for Normal
Operations. A description of how
programmatic controls, including
monitoring programs, would provide
assurance that design features and
procedures will enable the holder of an
OL or COL to comply with § 53.260.
(9) Design Features Supporting the
Protection of Plant Workers. A
description of the design features
required by § 53.430 to support the
holder of an OL or COL complying with
§ 53.270.
(10) Programmatic Controls for
Protection of Plant Workers. A
description of how programmatic
controls, including monitoring
programs, would provide assurance that
design features and procedures will
enable the holder of an OL or COL to
comply with § 53.270.
(11) Codes and Standards. A
description of generally accepted

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consensus codes and standards used to
design the design features, as required
by § 53.440(b).
(12) Materials. A description of the
materials used for SR and NSRSS SSCs
and a description of the qualification of
these materials for their service
conditions over the plant lifetime, as
required by § 53.440(c).
(13) Integrity Assessment Program. A
description of a design integrity
assessment program that addresses the
elements described in § 53.440(d).
(14) Safety and Security. Confirmation
that safety and security were considered
together in the design process, as
required by § 53.440(f).
(15) Criticality. Information
demonstrating how the applicant will
comply with requirements for criticality
accidents in § 53.440(m).
(16) Multi-unit Plants. For an
application for standard design
certification of a multi-unit commercial
nuclear plant, the possible operating
configurations of the reactor units,
including common systems, interface
requirements, and system interactions,
as required by § 53.440(i).
(17) SSC Classification. (i) The
classification of SSCs according to their
safety significance under § 53.460(a).
(ii) For SR and NSRSS SSCs, the
conditions under which they must
perform the safety functions required by
§ 53.230, including environmental
conditions.
(18) Probabilistic Risk Assessment. A
description of the probabilistic risk
assessment (PRA) required by
§ 53.450(a) and its results.
(19) Analyses. A description of the
analyses performed under § 53.450(b)
through (g) that includes the following
information:
(i) A description of the analysis of
LBEs and its results, as described in
§ 53.240. This analysis description
must—
(A) Address the elements in
§ 53.450(e) and (f); and
(B) Under § 53.460(c)—
(1) Describe any human actions that
are necessary to prevent or mitigate
LBEs;
(2) Describe how those human actions
are capable of being reliably performed
under the postulated environmental
conditions present; and
(3) Describe how those human actions
would be addressed by programs
established under subpart F of this part.
(ii)(A) A description of how SSCs
relied on to meet the safety criteria
defined in § 53.210 are protected against
or designed to withstand the effects of
external hazards under § 53.510.
(B) The information necessary to
demonstrate that the commercial

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nuclear plant complies with the
earthquake engineering criteria in
§ 53.480.
(iii) A description of the defense-indepth measures required by § 53.250.
(iv) A description of all plant
operating states where there is the
potential for the uncontrolled release of
radioactive material to the environment,
as required by § 53.450(b)(4).
(v) A description of the events that
challenge plant control and safety
systems whose failure could lead to an
undesirable end state and/or radioactive
material release, as required by
§ 53.450(b)(5).
(vi) A description of the analytical
codes used in modeling plant behavior
in analyses of LBEs and how these
codes are qualified for the range of
conditions for which they were used, as
required by § 53.450(d).
(vii) If not described in addressing
paragraph (5) of this section, the results
of other analyses required by
§ 53.450(g).
(20) Special Treatments. A
description of special treatments
established as required by § 53.460.
(21) Analytical Margins. A description
of any alternative criteria adopted to
demonstrate analytical margins
supporting operational flexibilities, if
applicable, as required by § 53.470.
(22) Quality Assurance. A description
of the QAP applied to the design of the
SSCs of the commercial nuclear plant,
as required by § 53.460(b). The
description of the QAP for a commercial
nuclear plant must include a discussion
of how the applicable requirements of
appendix B to part 50 of this chapter
were satisfied.
(23) Design Features and Controls to
Address the Minimization of
Contamination. The information
required by § 20.1406 of this chapter.
(24) Interface Requirements. (i) A
description analysis, and evaluation of
the interfaces between the standard
design and the balance of the
commercial nuclear plant that may
impact the ability of the plant to
demonstrate compliance with the
functional design criteria or the safety
criteria of subparts B and C of this part.
(ii) Confirmation that interface
requirements are verifiable through
inspections, testing, or analysis. These
requirements must be sufficiently
detailed to allow for completion of the
final safety analysis by license
applicants that reference the certified
design under this subpart. The method
to be used for verification of interface
requirements must be included as part
of the proposed ITAAC required by
§ 53.1241(a)(3).

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(iii) A representative conceptual
design for those portions of the plant for
which the application does not seek
certification to aid the NRC in its review
of the FSAR and to permit assessment
of the adequacy of the interface
requirements under paragraph (a)(24)(i)
of this section.
(25) Technical Qualifications. A
description of the technical
qualifications of the applicant to engage
in the proposed activities in accordance
with the regulations in this chapter.
(26) Technical Specifications.
Proposed technical specifications
prepared under § 53.710(a) for those
areas addressed by the design
certification.
(27) Role of personnel. Information to
address the following for the role of
personnel in ensuring safe operations:
(i) A description of how the human
factors engineering design requirements
of § 53.440(n)(1) are addressed;
(ii) A description of how the human
system interface design requirements of
§ 53.440(n)(2) are addressed;
(iii) A concept of operations that is of
sufficient scope and detail to address
the requirements of § 53.440(n)(3);
(iv) A functional requirements
analysis and function allocation that is
of sufficient scope and detail to address
the requirements of § 53.440(n)(4).
(b) [Reserved]

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§ 53.1241 Contents of applications for
standard design certifications; other
application content.

(a) In addition to the FSAR, the
application must also include the
following:
(1) Environmental report. An
environmental report as required by
§ 51.55 of this chapter.
(2) Availability Controls (if not
included in the FSAR). A description of
the controls on plant operations,
including availability controls, to
provide reasonable confidence that the
configurations and special treatments
for SR and NSRSS SSCs provide the
capabilities and reliabilities required to
demonstrate compliance with the safety
criteria of § 53.220, or more restrictive
alternative criteria adopted under
§ 53.470.
(3) Inspections, tests, analyses, and
acceptance criteria. The proposed
ITAAC that are necessary and sufficient
to provide reasonable assurance that, if
the inspections, tests, and analyses are
performed and the acceptance criteria
met, a facility that incorporates the
design certification has been
constructed and will be operated in
conformity with the design certification,
the provisions of the Act, and the
Commission’s rules and regulations.

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(4) Safeguards information. A
description of the program to protect
Safeguards Information against
unauthorized disclosure in accordance
with the requirements in §§ 73.21 and
73.22 of this chapter, as applicable.
(b) If there are SSCs of the plant
which required research and
development to confirm the adequacy of
their design, provide a report in the
application which documents the
resolution of any safety questions
associated with such SSCs.
(c) A description of how the
performance of each design feature has
been demonstrated capable of fulfilling
functional design criteria considering
interdependent effects through either
analysis, appropriate test programs,
prototype testing, operating experience,
or a combination thereof, in accordance
with § 53.440(a).
§ 53.1242

Review of applications.

(a) Standards for review of
applications. Applications filed under
this part will be reviewed for
compliance with the standards set out
in this part and 10 CFR parts 20, 51, and
73.
(b) Administrative review of
applications; hearings. (1) A standard
design certification is a rule that will be
issued under the provisions of subpart
H of 10 CFR part 2, as supplemented by
the provisions of this section. The
Commission must initiate the
rulemaking after an application has
been filed under § 53.1233 and must
specify the procedures to be used for the
rulemaking. The notice of proposed
rulemaking published in the Federal
Register must provide an opportunity
for the submission of comments on the
proposed design certification rule. If, at
the time a proposed design certification
rule is published in the Federal Register
under this paragraph, the Commission
decides that a legislative hearing should
be held, the information required by
§ 2.1502(c) of this chapter must be
included in the Federal Register
document for the proposed design
certification.
(2) Following the submission of
comments on the proposed design
certification rule, the Commission may,
at its discretion, hold a legislative
hearing under the procedures in subpart
O of part 2 of this chapter. The
Commission must publish a document
in the Federal Register of its decision to
hold a legislative hearing. The
document must contain the information
specified in § 2.1502(c) of this chapter
and specify whether the Commission or
a presiding officer will conduct the
legislative hearing.

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(3) Notwithstanding anything in
§ 2.390 of this chapter to the contrary,
proprietary information will be
protected in the same manner and to the
same extent as proprietary information
submitted in connection with
applications for licenses, provided that
the design certification will be
published in chapter I of this title.
(c) Reference to an issued operating
license or combined license. In those
cases where a design certification
application is preceded by the issuance
of an OL or custom COL for a
commercial nuclear plant that is
essentially the same as the standard
design for which certification is being
requested, the NRC review will follow
the processes for referencing a standard
design approval in § 53.1221, to the
extent practicable.
§ 53.1245 Referral to the Advisory
Committee on Reactor Safeguards.

The Commission must refer a copy of
the application to the ACRS. The ACRS
must report on those portions of the
application which concern safety.
§ 53.1248 Issuance of standard design
certification.

(a) After conducting a rulemaking
proceeding under § 53.1242 on an
application for a standard design
certification and receiving the report to
be submitted by the ACRS under
§ 53.1245, the Commission may issue a
standard design certification in the form
of a rule for the design that is the subject
of the application, if the Commission
determines that—
(1) The application demonstrates
compliance with the applicable
standards and requirements of the Act
and the Commission’s regulations;
(2) Notifications, if any, to other
agencies or bodies have been duly
made;
(3) There is reasonable assurance that
the standard design conforms with the
provisions of the Act and the
Commission’s regulations;
(4) The applicant is technically
qualified;
(5) The proposed ITAAC are
necessary and sufficient, within the
scope of the standard design, to provide
reasonable assurance that, if the
inspections, tests, and analyses are
performed and the acceptance criteria
met, the facility has been constructed
and will be operated in accordance with
the design certification, the provisions
of the Act, and the Commission’s
regulations;
(6) Issuance of the standard design
certification will not be inimical to the
common defense and security or to the
health and safety of the public;

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(7) The findings required by subpart
A of part 51 of this chapter have been
made; and
(8) The applicant has implemented
the QAP described or referenced in the
Safety Analysis Report.
(b) The design certification rule must
specify the site parameters, design
characteristics, and any additional
requirements and restrictions of the
design certification rule.
(c) After the Commission has adopted
a final design certification rule, the
applicant must not permit any
individual to have access to or any
facility to possess restricted data or
classified National Security Information
until the individual and/or facility has
been approved for access under the
provisions of 10 CFR parts 25 and/or 95,
as applicable.
§ 53.1251

Duration of certification.

(a) Except as provided in paragraph
(b) of this section, a standard design
certification issued under this subpart is
valid for 15 years from the effective date
of the rule.
(b) A standard design certification
continues to be valid beyond the date of
expiration in any proceeding on an
application for a COL or an OL under
this part that references the standard
design certification and is docketed
either before the date of expiration of
the certification, or, if a timely
application for renewal of the
certification has been filed, before the
Commission has determined whether to
renew the certification. A design
certification also continues to be valid
beyond the date of expiration in any
hearing held under § 53.1452 before
operation begins under a COL that
references the design certification.
(c) An applicant for a CP, OL, COL,
or ML under this part may, at its own
risk, reference in its application a design
for which a design certification
application has been docketed but not
granted.

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§ 53.1254

Application for renewal.

(a) Not less than 12 nor more than 36
months before the expiration of the
initial 15-year period, or any later
renewal period, any person may apply
for renewal of the certification. An
application for renewal must contain all
information necessary to bring up to
date the information and data contained
in the previous application. The
Commission will require, before
renewal of certification, that engineering
documents, such as analyses, drawings,
procurement specifications, or
construction and installation
specifications, be completed and
available for audit if the more detailed

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information is necessary for the
Commission to verify the information in
the application and make its safety
determination. Notice and comment
procedures must be used for a
rulemaking proceeding on the
application for renewal. The
Commission, in its discretion, may
require the use of additional procedures
in individual renewal proceedings.
(b) A design certification, either
original or renewed, for which a timely
application for renewal has been filed
remains in effect until the Commission
has determined whether to renew the
certification. If the certification is not
renewed, it continues to be valid in
certain proceedings under § 53.1251.
(c) The Commission must refer a copy
of the application for renewal to the
ACRS. The ACRS must report on those
portions of the application which
concern safety and must apply the
criteria set forth in § 53.1257.
§ 53.1257

Criteria for renewal.

(a) The Commission must issue a rule
granting the renewal if the design, either
as originally certified or as modified
during the rulemaking on the renewal,
complies with the Act and the
Commission’s regulations applicable
and in effect at the time the certification
was issued.
(b) The Commission may impose
other requirements if it determines
that—
(1) They are necessary for adequate
protection to public health and safety or
common defense and security;
(2) They are necessary for compliance
with the Commission’s regulations and
orders applicable and in effect at the
time the design certification was issued;
or
(3) There is a substantial increase in
overall protection of the public health
and safety or the common defense and
security to be derived from the new
requirements, and the direct and
indirect costs of implementing those
requirements are justified in view of this
increased protection.
(c) In addition, the applicant for
renewal may request an amendment to
the design certification. The
Commission must grant the amendment
request if it determines that the
amendment will comply with the Act
and the Commission’s regulations in
effect at the time of renewal. If the
amendment request entails such an
extensive change to the design
certification that an essentially new
standard design is being proposed, an
application for a design certification
must be filed in accordance with this
subpart.

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(d) Denial of renewal does not bar the
applicant, or another applicant, from
filing a new application for certification
of the design, which proposes design
changes that correct the deficiencies
cited in the denial of the renewal.
§ 53.1260

Duration of renewal.

Each renewal of certification for a
standard design will be for not less than
10, nor more than 15 years.
§ 53.1263 Finality of standard design
certifications.

(a)(1) While a standard design
certification rule is in effect under
§§ 53.1251 or 53.1260, the Commission
may not modify, rescind, or impose new
requirements on the certification
information, whether on its own
motion, or in response to a petition from
any person, unless the Commission
determines in a rulemaking that the
change—
(i) Is necessary either to bring the
certification information or the
referencing plants into compliance with
the Commission’s regulations applicable
and in effect at the time the certification
was issued;
(ii) Is necessary to provide adequate
protection of the public health and
safety or the common defense and
security;
(iii) Reduces unnecessary regulatory
burden and maintains protection to
public health and safety and the
common defense and security;
(iv) Provides the detailed design
information to be verified under those
ITAAC that are directed at certification
information (i.e., design acceptance
criteria);
(v) Is necessary to correct material
errors in the certification information;
(vi) Substantially increases overall
safety, reliability, or security of facility
design, construction, or operation, and
the direct and indirect costs of
implementation of the rule change are
justified in view of this increased safety,
reliability, or security; or
(vii) Contributes to increased
standardization of the certification
information.
(2)(i) In a rulemaking under
§ 53.1263(a)(1), except for
§ 53.1263(a)(1)(ii), the Commission will
give consideration to whether the
benefits justify the costs for plants that
are already licensed or for which an
application for a permit or license is
under consideration.
(ii) The rulemaking procedures for
changes under § 53.1263(a)(1) must
provide for notice and opportunity for
public comment.
(3) Any modification the NRC
imposes on a design certification rule

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under paragraph (a)(1) of this section
will be applied to all plants referencing
the certified design, except those to
which the modification has been
rendered technically irrelevant by
action taken under paragraphs (a)(4) or
(b) of this section.
(4) The Commission may not impose
new requirements by plant-specific
order on any part of the design of a
specific plant referencing the design
certification rule if that part was
approved in the design certification
while a design certification rule is in
effect under § 53.1248, unless—
(i) A modification is necessary to
secure compliance with the
Commission’s regulations applicable
and in effect at the time the certification
was issued, or to assure adequate
protection of the public health and
safety or the common defense and
security; and
(ii) Special circumstances as defined
in § 53.080 are present. In addition to
the factors listed in § 53.080, the
Commission must consider whether the
special circumstances which § 53.080
requires to be present outweigh any
decrease in safety that may result from
the reduction in standardization caused
by the plant-specific order.
(5) Except as provided in § 2.335 of
this chapter, in making the findings
required for issuance of a COL, CP, OL,
or ML, or for any hearing under
§ 53.1452, the Commission must treat as
resolved those matters resolved in
connection with the issuance or renewal
of a design certification rule.
(b) An applicant who references a
design certification rule may request an
exemption from one or more elements of
the certification information. The
Commission may grant such a request
only if it determines that the exemption
will comply with the requirements of
§ 53.080. In addition to the factors listed
in § 53.080, the Commission must
consider whether the special
circumstances that § 53.080 requires to
be present outweigh any decrease in
safety that may result from the
reduction in standardization caused by
the exemption. The granting of an
exemption on request of an applicant is
subject to litigation in the same manner
as other issues in the OL or COL
hearing.
(c) The Commission will require,
before granting a CP, COL, OL, or ML
that references a design certification
rule, that information normally
contained in engineering documents,
such as analyses, drawings,
procurement specifications, or
construction and installation
specifications, be completed and
available for audit if the more detailed

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information is necessary for the
Commission to verify the information in
the application and make its safety
determination, including the
determination that the application is
consistent with the certification
information. This information may be
acquired by appropriate arrangements
with the design certification applicant.
§ 53.1270

Manufacturing licenses.

Sections 53.1270 through 53.1295 set
out the requirements and procedures
applicable to Commission issuance of a
license under this part authorizing
manufacture of manufactured reactors to
be installed at sites not identified in the
ML application.
§ 53.1273

Filing of applications.

Any person, except one excluded by
§ 53.1118, may file an application for an
ML under this part with the Director,
Office of Nuclear Reactor Regulation.
§ 53.1276 Contents of applications for
manufacturing licenses; general
information.

Each application for an ML must
include the information contained in
§ 53.1109(a) through (e), and (j).
§ 53.1279 Contents of applications for
manufacturing licenses; technical
information.

(a) Final Safety Analysis Report-siting
and design. The application must
include an FSAR containing the
information set forth below, with a level
of design information sufficient to
enable the Commission to judge the
applicant’s proposed means of ensuring
that the manufacturing conforms to the
design and to reach a final conclusion
on all safety questions associated with
the design, permit the preparation of
construction and installation
specifications by an applicant who
seeks to use the manufactured reactor,
and permit the preparation of
acceptance and inspection requirements
by the NRC. The application must
include the following information:
(1) Site parameters. The site
parameters postulated for the design
under this part, including the designbasis external hazard levels for the
relevant external hazards, and an
analysis and evaluation of the design in
terms of those site parameters.
(2) Design information. Except as
specified in this paragraph, the design
information equivalent to that required
for a standard design certification as
defined in § 53.1239(a)(2) through (27)
for those portions of a commercial
nuclear plant included in the
manufactured reactor.
(3) Quality assurance program. A
description of the QAP applied to the

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design, and to be applied to the
fabrication and testing of the SSCs of the
manufactured reactor under
§ 53.620(a)(6), including a discussion of
how the applicable requirements of
appendix B to part 50 of this chapter
will be satisfied;
(4) Conceptual designs.
Representative conceptual designs for
one or more commercial nuclear plants
using the manufactured reactor;
(5) Operating configurations. If
multiple manufactured reactors may be
installed at a commercial nuclear plant,
a description of the possible operating
configurations, including common
systems, interface requirements, and
system interactions. The final safety
analysis must also account for
differences among the possible
configurations, including any
restrictions that will be necessary
during the construction and startup of a
given manufactured reactor to ensure
the safe operation of any commercial
nuclear reactor already operating;
(6) Interface requirements. (i) The
interface requirements between the
manufactured reactor and the remaining
portions of the commercial nuclear
plant or connections to other facilities
outside of the commercial nuclear plant.
(ii) Confirmation that interface
requirements are verifiable through
inspections, testing, or analysis. These
requirements must be sufficiently
detailed to allow for completion of the
final safety analysis by license
applicants that reference the
manufactured reactor manufactured
under this subpart. Applicants for a
COL under this part will need to verify
the interface requirements at the
installation site. The method to be used
for verification of interface requirements
must be included as part of the
proposed ITAAC required by
§ 53.1282(a).
(iii) Information to support
development of radiation monitoring
programs required under subpart F of
this part by an applicant for a COL,
including potential pathways for
radionuclides produced within the
manufactured reactor to enter
interfacing systems.
(b) Final Safety Analysis Report—
manufacturing information. The FSAR
must include the following information
related to the manufacturing processes,
organization, controls, and inspections:
(1) A description, including
references to generally accepted
consensus codes and standards, of the
processes that will be used to procure,
fabricate, and assemble components that
make up the manufactured reactor. The
description should clearly define which
activities are proposed to be within the

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scope of the ML and those, such as the
making of a component to be procured
from a separate company for installation
in the manufactured reactor, that are not
considered to be within the scope of the
ML;
(2) A description of the organizational
and management structure singularly
responsible for direction of design and
manufacture of the manufactured
reactor. The information should include
a description of the management plans,
technical qualifications, and controls in
place to demonstrate compliance with
the requirements of § 53.620;
(3) A description of the inspections
and tests to be performed as part of the
manufacturing process, including the
inspection of procured components,
inspection and testing of fabrication
processes such as the molding, welding,
or coating of components, and
inspections and testing of the assembled
manufactured reactor or portions of the
manufactured reactor;
(4) A description of the fitness-forduty program required by part 26 of this
chapter and its implementation.
(c) Deployment of the completed
manufactured reactor. The application
must include the following information
related to the deployment of a
manufactured reactor:
(1) Procedures governing the
preparation of the manufactured reactor
or portions of the manufactured reactor
for shipping to the site where it is to be
operated; the conduct of shipping; and
verifying the condition of the shipped
items upon receipt at the site;
(2) Details of the interaction of the
design, manufacture, and installation of
a manufactured reactor within the
applicant’s organization and the manner
by which the applicant will ensure close
integration between the designer,
contractors, and any facility in which
the manufactured reactor is to be
installed;
(3) A description of the measures to
be used for the control of interfaces,
including the consideration of key site
parameters, between the holder of the
ML and the holder of the COL for the
commercial nuclear plant at which the
manufactured reactor is to be installed.
(d) Special considerations for factory
fueling. In addition to the above
paragraphs, an application for an ML for
a manufactured reactor that will be
fueled at the factory under a 10 CFR part
70 license must include the following
information related to loading fuel and
the required independent physical
mechanisms to prevent criticality and to
otherwise provide assurance that the
fueled manufactured reactor can be
successfully transported, installed, and
operated at a site for which the

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Commission has issued a COL that
authorizes construction and operation of
a commercial nuclear plant using the
manufactured reactor:
(1) A description of the procedures
used during the fueling of the
manufactured reactor that ensure that
the configuration of fuel within the
fueled manufactured reactor is
consistent with the design and analyses
supporting operation of the
manufactured reactor under the COL at
the place of operation. The description
may reference the applicable 10 CFR
part 70 application and other sections of
the Safety Analysis Report supporting
the ML license application.
(i) The application must describe the
measures taken for in-factory
inspections and non-nuclear testing
performed to ensure that the
configuration of fuel within the fueled
manufactured reactor is consistent with
the design and analyses supporting
operation of the manufactured reactor
under the COL at the place of operation.
(ii) The application must describe the
design features included in the
manufactured reactor to prevent
criticality, including at least two
independent mechanisms each of which
is sufficient to prevent criticality, the
associated functional design criteria
applied to those design features, and the
physical and programmatic controls
implemented during manufacturing,
storage, and transport that are credited
to assure the features function as
designed when subject to potential
hazards and human errors. The
descriptions must include how those
measures will be controlled during
installation under the ML and removal
under the COL at the place of operation.
(2) A description of the procedures
governing the transfer of responsibilities
for the fueled manufactured reactor
from the holder of the ML to the holder
of the COL for the installation site.
(3) If available at the time of filing the
ML application or, if not available at the
time of filing the ML application,
submitted as an amendment to the ML
or ML application at the time of filing
the Part 70 application, a description of
the programs needed to demonstrate
compliance with the requirements of
§ 53.620(d) and 10 CFR parts 70, 71, and
73 for the receipt, storage, and loading
of SNM into a manufactured reactor and
the transport of the fueled manufactured
reactor to a site for which the
Commission has issued a COL that
authorizes construction and operation of
a commercial nuclear plant using the
manufactured reactor, including the
following.
(i) A physical security program in
accordance with § 53.620(d)(2)(i).

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(ii) A cybersecurity program in
accordance with § 53.620(d)(2)(i).
§ 53.1282 Contents of applications for
manufacturing licenses; other application
content.

(a) Inspections, tests, analyses, and
acceptance criteria. (1) The application
must contain proposed inspections,
tests, and analyses that the COL holder
must perform, and the acceptance
criteria that are necessary and sufficient
to provide reasonable assurance that, if
the inspections, tests, and analyses are
performed and the acceptance criteria
met:
(i) The reactor has been manufactured
in conformity with the ML, the
provisions of the Act, and the
Commission’s rules and regulations; and
(ii) The manufactured reactor will be
operated in conformity with the
approved design and any license
authorizing operation of the
manufactured reactor.
(2) If the application references a
standard design certification, the ITAAC
contained in the certified design must
apply to those portions of the facility
design that are covered by the design
certification.
(3) If the application references a
standard design certification, the
application may include a notification
that a required inspection, test, or
analysis in the design certification
ITAAC has been successfully completed
and that the corresponding acceptance
criterion has been met. The Federal
Register notification required by
§ 53.1285 must indicate that the
application includes this notification.
(b) Environmental report. (1) The
application must contain an
environmental report as required by
§ 51.54 of this chapter.
(2) If the ML application references a
standard design certification, the
environmental report need not contain a
discussion of severe accident mitigation
design alternatives for the manufactured
reactor as used in a commercial nuclear
plant.
(c) Safeguards information. The
application must contain a description
of the program to protect safeguards
information against unauthorized
disclosure in accordance with the
requirements in §§ 73.21 and 73.22 of
this chapter, as applicable.
(d) Performance demonstration. A
description of how the performance of
each design feature has been
demonstrated capable of fulfilling
functional design criteria considering
interdependent effects through either
analysis, appropriate test programs,
prototype testing, operating experience,

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or a combination thereof, in accordance
with § 53.440(a).
§ 53.1285

Review of applications.

(a) Standards for review of
applications. Applications for MLs
under this part will be reviewed
according to the applicable standards
set out in this subpart as well as
applicable standards in this part and 10
CFR parts 20, 25, 26, 51, 70, 71, 73, and
75.
(b) Administrative review of
applications, hearings. A proceeding on
an ML is subject to all applicable
procedural requirements contained in
10 CFR part 2, including the
requirements for docketing in
§ 2.101(a)(1) through (4) of this chapter,
and the requirements for issuance of a
notice of proposed action in § 2.105 of
this chapter, provided, however, that the
designated sections may not be
construed to require that the
environmental report or draft or final
environmental impact statement include
an assessment of the benefits of
constructing and/or operating the
manufactured reactor or an evaluation
of alternative energy sources. All
hearings on MLs are governed by the
hearing procedures contained in 10 CFR
part 2, subparts C, E, G, L, and N.
§ 53.1286 Referral to the Advisory
Committee on Reactor Safeguards.

The Commission must refer a copy of
the application to the ACRS. The ACRS
must report on those portions of the
application which concern safety.

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§ 53.1287
licenses.

Issuance of manufacturing

(a) After completing any hearing
under § 53.1285(b), and receiving the
report submitted by the ACRS, the
Commission may issue an ML if the
Commission finds that—
(1) Applicable standards and
requirements of the Act and the
Commission’s regulations have been
met;
(2) There is reasonable assurance that
the manufactured reactor will be
manufactured, and can be transported,
incorporated into a commercial nuclear
plant, and operated in conformity with
the ML, the provision of the Act, and
the Commission’s regulations;
(3) The proposed manufactured
reactor can be incorporated into a
commercial nuclear plant and operated
at sites having characteristics that fall
within the site parameters postulated for
the design of the manufactured reactors
without undue risk to the health and
safety of the public;
(4) The applicant is technically
qualified to design and manufacture the
proposed manufactured reactor;

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(5) The proposed ITAAC are
necessary and sufficient, within the
scope of the ML, to provide reasonable
assurance that the manufactured reactor
has been manufactured and will be
operated in conformity with the license,
the provisions of the Act, and the
Commission’s regulations;
(6) The issuance of a license to the
applicant will not be inimical to the
common defense and security or to the
health and safety of the public; and
(7) The findings required by subpart
A of 10 CFR part 51 have been made.
(b) Each ML issued under this subpart
must specify—
(1) Terms and conditions as the
Commission deems necessary and
appropriate;
(2) Technical specifications for
operation of the manufactured reactor,
as the Commission deems necessary and
appropriate;
(3) Site parameters and design
characteristics for the manufactured
reactor;
(4) The interface requirements to be
met by the site-specific elements of the
facility, such as the energy conversions
systems and ultimate heat sink, not
within the scope of the manufactured
reactor; and
(5) The entity with design authority
for the manufactured reactor covered by
the license.
§ 53.1288
licenses.

Finality of manufacturing

(a)(1) Notwithstanding any provision
in § 53.1590, during the term of an ML
issued under this part the Commission
may not modify, rescind, or impose new
requirements on the design of the
manufactured reactor, or the
requirements for the manufacture of the
manufactured reactor, unless the
Commission determines that a
modification is necessary to bring the
design of the reactor or its manufacture
into compliance with the Commission’s
requirements applicable and in effect at
the time the ML was issued, or to
provide reasonable assurance of
adequate protection to public health and
safety or common defense and security.
(2) Any modification to the design of
a manufactured reactor that is imposed
by the Commission under paragraph
(a)(1) of this section will be applied to
all manufactured reactors manufactured
under the license, including those that
have already been transported and sited,
except those manufactured reactors to
which the modification has been
rendered technically irrelevant by
action taken under § 53.1530 or
paragraph (b) of this section.
(3) In making the findings required
under this part for issuance of a COL,

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in any hearing under § 53.1452, or in
any enforcement hearing other than one
initiated by the Commission under
paragraph (a)(1) of this section, for
which a manufactured reactor
manufactured under this subpart is
referenced or used, the Commission
must treat as resolved those matters
resolved in the proceeding on the
application for issuance or renewal of
the ML, including the adequacy of
design of the manufactured reactor, the
costs and benefits of severe accident
mitigation design alternatives, and the
bases for not incorporating severe
accident mitigation design alternatives
into the design of the manufactured
reactor to be manufactured.
(b) An applicant who references or
uses a manufactured reactor
manufactured under an ML under this
part may include in the application a
request for a departure from the design
characteristics, site parameters, terms
and conditions, or approved design of
the manufactured reactor. The
Commission may grant a request only if
it determines that the departure will
comply with the requirements of
§ 53.080, and that the special
circumstances outweigh any decrease in
safety that may result from the
reduction in standardization caused by
the departure. The granting of a
departure on request of an applicant is
subject to litigation in the same manner
as other issues in the COL hearing.
§ 53.1291
licenses.

Duration of manufacturing

An ML issued under this part is valid
for not less than 5, nor more than 15
years from the date of issuance. Upon
expiration of the ML, the manufacture of
any uncompleted manufactured reactors
must cease unless a timely application
for renewal has been docketed with the
NRC.
§ 53.1293
licenses.

Transfer of manufacturing

An ML may be transferred under
§ 53.1570.
§ 53.1295
licenses.

Renewal of manufacturing

(a)(1) Not less than 12 months, nor
more than 5 years before the expiration
of the ML, or any later renewal period,
the holder of the ML issued under this
part may apply for a renewal of the
license. An application for renewal
must contain all information necessary
to bring up to date the information and
data contained in the previous
application.
(2) The filing of an application for a
renewed license must be in accordance
with subpart A of 10 CFR part 2 and
§ 53.1100.

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(3) An ML issued under this part,
either original or renewed, for which a
timely application for renewal has been
filed, remains in effect until the
Commission has made a final
determination on the renewal
application, provided, however, that the
holder of an ML may not begin
manufacture of a manufactured reactor
less than 6 months before the expiration
of the license.
(4) Any person whose interest may be
affected by renewal of the license may
request a hearing on the application for
renewal. The request for a hearing must
comply with § 2.309 of this chapter. If
a hearing is granted, notice of the
hearing will be published in accordance
with § 2.104 of this chapter.
(5) The Commission must refer a copy
of the application for renewal to the
ACRS. The ACRS must report on those
portions of the application which
concern safety.
(b) The Commission may grant the
renewal if the Commission
determines—
(1) The ML complies with the Act and
the Commission’s regulations and
orders applicable and in effect at the
time the ML was originally issued; and
(2) Any new requirements the
Commission may wish to impose are—
(i) Necessary for adequate protection
to public health and safety or common
defense and security;
(ii) Necessary for compliance with the
Commission’s regulations and orders
applicable and in effect at the time the
ML was originally issued; or
(iii) A substantial increase in overall
protection of the public health and
safety or the common defense and
security to be derived from the new
requirements, and the direct and
indirect costs of implementation of
those requirements are justified in view
of this increased protection.
(c) A renewed ML may be issued for
a term of not less than 5, nor more than
15 years, plus any remaining years on
the ML then in effect before renewal.
The renewed license must be subject to
the requirements of § 53.1288.

An application for a CP must include
the information required by § 53.1109
and the following information:
(a) Information sufficient to
demonstrate to the Commission the
financial qualification of the applicant
to carry out, under the regulations in
this chapter, the activities for which the
permit is sought. As applicable, the
following should be provided:
(1) The information that demonstrates
that the applicant possesses or has
reasonable assurance of obtaining the
funds necessary to cover estimated
construction costs and related fuel cycle
costs, including estimates of the total
construction costs and related fuel cycle
costs of the facility and must indicate
the source(s) of funds to cover these
costs.
(2) Each application for a CP
submitted by a newly formed entity
organized for the primary purpose of
constructing and operating a facility
must also include information showing:
(i) The legal and financial
relationships the entity has or proposes
to have with its stockholders or owners;
(ii) The stockholders’ or owners’
financial ability to meet any contractual
obligation to the entity that they have
incurred or proposed to incur; and
(iii) Any other information considered
necessary by the Commission to enable
it to determine the applicant’s financial
qualification; and
(3) The Commission may request an
established entity or newly-formed
entity to submit additional or more
detailed information respecting its
financial arrangements and status of
funds if the Commission considers this
information appropriate. This may
include information regarding an
applicant’s ability to continue the
conduct of the activities authorized by
the CP and to decommission the facility.
(b) If the applicant proposes to
construct or alter a facility, the
application must state the earliest and
latest dates for completion of the
construction or alteration.

Construction permits.

§ 53.1309 Contents of applications for
construction permits; technical information.

Sections 53.1300 through 53.1348 set
out the requirements and procedures
applicable to Commission issuance of a
CP for commercial nuclear plants. A CP
for the construction of a commercial
nuclear plant under this part will be
issued before the issuance of an OL if
the application is otherwise acceptable
and will be converted upon completion
of the facility and Commission action,
into an OL as provided under
§§ 53.1360 through 53.1405.

The application must contain a
Preliminary Safety Analysis Report
(PSAR) that describes the facility and
the limits on its operation and presents
a preliminary safety analysis of the SSCs
of the facility as a whole. The PSAR
must include the following information,
at a level of detail sufficient to enable
the Commission to reach a conclusion
on safety matters that must be resolved
by the Commission before issuance of a
CP:

§ 53.1300

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§ 53.1306 Contents of applications for
construction permits; general information.

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(a)(1) Site information. An application
for a CP for a commercial nuclear
reactor must include the site
information equivalent to that required
for an early site permit in
§ 53.1146(a)(1)(iv) through (x).
(2) Design information. Except as
specified in this paragraph, an
application for a CP for a commercial
nuclear plant must include the design
information equivalent to that required
for a standard design certification as
defined in § 53.1239(a)(2) through (27).
(i) Quality assurance program. A
description of the QAP to be applied to
the design, fabrication, construction,
and testing of the SSCs of the facility
under § 53.610(a)(6), including a
discussion of how the requirements of
appendix B to part 50 of this chapter
will be satisfied.
(ii) Preliminary design information.
The information provided in the
application may include some aspects of
the design that are not fully developed,
and the information is therefore
preliminary. The completed design,
including any changes during
construction, must be described in the
FSAR required in § 53.1369 that
supports an application for an OL.
(iii) Planned research or testing.
Descriptions of how design features and
related functional design criteria will
fulfill the safety criteria in subpart B, or
more restrictive alternative criteria
adopted under § 53.470, and how that
has been or will be demonstrated
through either analysis, appropriate test
programs, experience, or a combination
thereof. Where any design feature has
not been fully developed or
demonstrated to fulfill the functional
design criteria at the time of an
application for a CP, the applicant must
provide a plan for future analysis,
research and development, test
programs, gathering of experience, or a
combination thereof to provide
reasonable confidence that the required
demonstration will be available for an
application for an OL.
(iv) Programmatic controls.
Descriptions of the programmatic
controls may include those to be
provided in the FSAR or other licensing
basis documents because they are
necessary to achieve and maintain the
reliability and capability of SSCs relied
upon to demonstrate compliance with
the established safety criteria and
functional design criteria required in
subpart B, and to maintain consistency
with analyses required by § 53.450.
(3) Technical qualifications. A
description of the technical
qualifications of the applicant to engage
in the proposed activities under the
regulations in this chapter.

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(4) Emergency preparedness. A
description of the applicant’s
preliminary plans for coping with
emergencies based on:
(i) Except as provided in paragraph
(a)(4)(ii) of this section, the
requirements in appendix E to part 50.
(ii) For a commercial nuclear plant
consisting of either small modular
reactors or non-light-water reactors, the
requirements in either § 50.160 or
appendix E to part 50.
(5) Physical security. A report that
provides a preliminary description of
how the site characteristics support the
development of adequate security plans
and measures consistent with the
requirements in § 53.540.
(6) Fitness-for-duty program. A
description of the fitness-for-duty (FFD)
program required by 10 CFR part 26 and
its implementation.
(b) A description of the program to
protect Safeguards Information against
unauthorized disclosure in accordance
with the requirements in §§ 73.21 and
73.22 of this chapter, as applicable.

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§ 53.1312 Contents of applications for
construction permits; other application
content.

(a) In addition to the PSAR, the
application must include the following:
(1) An environmental report either
under § 51.50(a) of this chapter if an
LWA under § 53.1130 is not requested
in conjunction with the CP application,
or under §§ 51.49 and 51.50(a) of this
chapter if an LWA is requested in
conjunction with the CP application; or
(2) If the applicant wishes to request
that an LWA under § 53.1130 be issued
before issuance of the CP, the
information otherwise required by
§ 53.1130, in accordance with either
§ 2.101(a)(1) through (a)(5), or
§ 2.101(a)(9) of this chapter.
(b) If the CP application references an
early site permit, standard design
approval, or standard design
certification issued under this part, then
the following requirements apply:
(1) The PSAR need not contain
information or analyses submitted to the
Commission in connection with the
referenced NRC approval, permit, or
certification, provided, however, that
the PSAR incorporates the material by
reference and confirms that the site and
design of the facility falls within
parameter values postulated in the
referenced NRC approval, permit, or
certification.
(2) The PSAR must provide a means
to demonstrate that all terms and
conditions that have been included in
the referenced NRC approval, permit, or
certification will be satisfied by the date
of issuance of the OL, as appropriate. If

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the PSAR does not demonstrate that
each site characteristic falls within the
corresponding postulated site parameter
and each design characteristic of the
facility falls within the corresponding
postulated design parameter, the
application must justify a departure,
variance, or exemption from the
referenced NRC approval, license, or
certification in regard to that particular
site or design characteristic in
compliance with the requirements of
this part.
(3) If a referenced early site permit
approves complete and integrated
emergency plans, or major features of
emergency plans, then the PSAR must
include any new or additional
information that updates and corrects
the information that was provided
under § 53.1146(b)(2) and discuss
whether the new or additional
information materially changes the
bases for compliance with the
applicable requirements.
§ 53.1315

Review of applications.

(a) Standards for review of
applications. Applications filed under
this part will be reviewed according to
the standards set out in this part and 10
CFR parts 20, 51, 73, and 140.
(b) Administrative review of
applications; hearings. A proceeding on
a CP application is subject to all
applicable procedural requirements
contained in 10 CFR part 2, including
the requirements for docketing (§ 2.101
of this chapter) and issuance of a notice
of hearing (§ 2.104 of this chapter). All
hearings on CP applications are
governed by the procedures contained
in 10 CFR part 2.
§ 53.1318 Finality of referenced NRC
approvals, permits, and certifications.

If the application for a CP under this
part references an early site permit,
standard design approval, or standard
design certification, the scope and
nature of matters resolved for the
application are governed by the relevant
provisions addressing finality, including
§§ 53.1188, 53.1221, and 53.1263.
§ 53.1324 Referral to the Advisory
Committee on Reactor Safeguards.

The Commission must refer a copy of
the application to the ACRS. The ACRS
must report on those portions of the
application that concern safety and
must apply the standards referenced in
§ 53.1315, in accordance with the
finality provisions in § 53.1318.
§ 53.1327 Authorization to conduct limited
work authorization activities.

(a) If the application does not
reference an early site permit which
authorizes the holder to perform the

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activities under § 53.1130, the applicant
may not perform those activities
without obtaining the separate
authorization required by § 53.1130.
Authorization may be granted only after
the presiding officer in the proceeding
on the application has made the
findings and determination required by
§ 53.1130(b)(1)(ii) and (iv), and the
Director, Office of Nuclear Reactor
Regulation makes the determination
required by § 53.1130(b)(1)(iii).
(b) If, after an applicant has performed
the activities permitted by paragraph (a)
of this section, the application for the
CP is withdrawn or denied, then the
applicant must implement an approved
site redress plan.
§ 53.1330 Exemptions, departures, and
variances.

(a) Applicants for a CP under this
part, or any amendment to a CP, may
include in the application a request for
an exemption from one or more of the
Commission’s regulations. The
Commission may grant a request if it
determines that the exemption complies
with § 53.080.
(b) An applicant for a CP who has
filed an application referencing an NRC
approval, permit, or certification issued
under this part may include in the
application a request for exemptions,
departures, or variances related to the
subject referenced NRC approval,
permit, or certification. In determining
whether to grant the departure,
variance, or exemption, the Commission
must apply the same technically
relevant criteria as were applicable to
the application for the original or
renewed approval, license, or
certification.
§ 53.1333
permits.

Issuance of construction

(a) After conducting a hearing in
accordance with § 53.1315 and receiving
the report submitted by the ACRS, the
Commission may issue a CP only if the
Commission finds that—
(1) The applicant has described the
proposed design of the facility and has
identified the major features or
components incorporated therein for the
protection of the health and safety of the
public;
(2) Such further technical or design
information as may be required to
complete the safety analysis, and which
can reasonably be left for later
consideration, will be supplied in the
FSAR;
(3) Safety features or components, if
any, that require research and
development have been described by
the applicant and the applicant has
identified, and there will be conducted,

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a research and development program
reasonably designed to resolve any
safety questions associated with such
features or components; and
(4) On the basis of the foregoing, there
is reasonable assurance of the
following—
(i) Such safety questions will be
satisfactorily resolved at or before the
latest date stated in the application for
completion of construction of the
proposed facility; and
(ii) Taking into consideration the site
criteria contained subpart D to this part,
the proposed facility can be constructed
and operated at the proposed location
without undue risk to the health and
safety of the public.
(b) A CP must contain the terms and
conditions for the permit, as the
Commission deems necessary and
appropriate. The Commission may, in
its discretion, incorporate in any CP
provisions requiring the applicant to
furnish periodic reports of the progress
and results of research and development
programs designed to resolve safety
questions.
§ 53.1336

Finality of construction permits.

Notwithstanding any provision in
§ 53.1590, a CP constitutes an
authorization to proceed with
construction but does not constitute
Commission approval of the safety of
any design feature or specification
unless the applicant specifically
requests such approval and such
approval is incorporated in the permit.
The applicant, at its option, may request
such approvals in the CP or by
amendment to the CP. If approved by
the NRC and included in the permit, the
NRC will consider modifications to the
approved design features or
specifications in accordance with
§ 53.1590.

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§ 53.1342
permits.

Duration of construction

(a) A CP will state the earliest and
latest dates for completion of
construction or alteration of the facility,
not to exceed 40 years from date of
issuance.
(b) If the proposed construction or
alteration of the facility is not
completed by the latest completion date,
the CP shall expire, and all rights are
forfeited. However, upon good cause
shown, the Commission will extend the
completion date for a reasonable period
of time. The Commission will recognize,
among other things, developmental
problems attributed to the experimental
nature of the facility or fire, flood
explosion, strike, sabotage, domestic
violence, enemy action an act of the
elements, and other acts beyond the

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control of the permit holder, as a basis
for extending the completion date.
§ 53.1345
permits.

Transfer of construction

A CP may be transferred under
§ 53.1570.
§ 53.1348
permits.

Termination of construction

When a permit holder has determined
to permanently cease construction, the
holder must, within 30 days, submit a
written certification to the NRC.
§ 53.1360

Operating licenses.

Sections 53.1360 through 53.1405 set
out the requirements and procedures
applicable to Commission issuance of
an OL for a nuclear power facility.
§ 53.1366 Contents of applications for
operating licenses; general information.

An application for an OL must
include the information required by
§ 53.1109 and the following
information:
(a) Except for an electric utility
applicant, information sufficient to
demonstrate to the Commission the
financial qualification of the applicant
to carry out, in accordance with the
regulations in this chapter, the activities
for which the license is sought. As
applicable, the following should be
provided:
(1) The applicant must submit
information that demonstrates the
applicant possesses or has reasonable
assurance of obtaining the funds
necessary to cover estimated operation
costs for the period of the license. The
applicant must submit estimates for
total annual operating costs for each of
the first 5 years of operation of the
facility. The applicant must also
indicate the source(s) of funds to cover
these costs.
(2) Each application for an OL
submitted by a newly-formed entity
organized for the primary purpose of
operating the facility must also include
information showing—
(i) The legal and financial
relationships the entity has or proposes
to have with its stockholders or owners;
(ii) The stockholders’ or owners’
financial ability to meet any contractual
obligation to the entity which they have
incurred or proposed to incur; and
(iii) Any other information considered
necessary by the Commission to enable
it to determine the applicant’s financial
qualification.
(3) The Commission may request an
established entity or newly formed
entity to submit additional or more
detailed information respecting its
financial arrangements and status of
funds if the Commission considers this

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information appropriate. This may
include information regarding a
licensee’s ability to continue the
conduct of the activities authorized by
the license and to decommission the
facility.
(b) The application must include
information in the form of a report, as
described in subpart G, indicating how
reasonable assurance will be provided
that funds will be available to
decommission the facility, including a
copy of the financial instrument
obtained to satisfy the requirements of
§ 53.1040.
§ 53.1369 Contents of applications for
operating licenses; technical information.

Final Safety Analysis Report. The
application must contain an FSAR that
describes the facility and the limits on
its operation and presents a safety
analysis of the SSCs of the facility as a
whole. The FSAR must include the
following information, at a level of
detail sufficient to enable the
Commission to reach a final conclusion
on all safety matters that must be
resolved by the Commission before
issuance of an OL. The FSAR must
include the following information:
(a) Site information. An application
for an OL for a commercial nuclear
reactor must include the site
information equivalent to that required
for an early site permit in
§ 53.1146(a)(1)(iv) through (x), including
all current information, such as the
results of environmental and
meteorological monitoring programs,
which has been developed since
issuance of the CP, relating to site
evaluation factors identified in this part.
(b) Design information. Except as
specified in this paragraph, an FSAR for
an OL for a commercial nuclear plant
must include the final design
information equivalent to that required
for a standard design certification as
defined in § 53.1239(a)(2) through (7),
(a)(9), and (a)(11) through (a)(27).
(1) The completed design, including
any changes during construction, must
be described.
(2) Where any design feature had not
been fully developed or demonstrated at
the time of application for the CP, the
applicant must provide the analysis,
research and development, test
programs, gathering of experience, or a
combination thereof to provide the
required demonstration to fulfill the
functional design criteria.
(c) Technical qualifications. A
description of the technical
qualifications of the applicant to engage
in the proposed activities in accordance
with the regulations in this chapter.

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(d) Integrity assessment program. A
description of an Integrity Assessment
Program that addresses the elements
described in § 53.870.
(e) Safeguards information. A
description of the program to protect
Safeguards Information against
unauthorized disclosure in accordance
with the requirements in §§ 73.21 and
73.22 of this chapter, as applicable.
(f) Emergency response facility or
facilities. Description of location and
capabilities to be established for
command and control, support, and
coordination of onsite and offsite, as
applicable, functions during reactor
accident conditions.
(g) Role of personnel. (1) A
description of the completed
assessments related to the role of
personnel in ensuring safe operations
considering the analyses required by
§ 53.730. These assessments must
include the following:
(i) Human factors engineering design
requirements of § 53.730(a);
(ii) Human system interface design
requirements of § 53.730(b);
(iii) Concept of operations of
§ 53.730(c);
(iv) Functional requirements analysis
and function allocation of § 53.730(d);
(2) A description of the program to be
used for evaluating and applying
operating experience as required by
§ 53.730(e);
(3) A staffing plan and supporting
analyses as required by § 53.730(f).
(h) Training, examination, and
proficiency programs. (1) A description
of the training, examination, and
proficiency programs required by
§ 53.730(g);
(2) A description of the training
programs required by § 53.830.
(i) Emergency plan. Emergency plans
complying with the requirements of
§ 53.855.
(1) Include all emergency plan
certifications, as applicable, that have
been obtained from the State, local, and
participating Tribal governmental
agencies with emergency planning
responsibilities that are wholly or
partially within the EPZ plume
exposure pathway. These certifications
must state that—
(i) The proposed emergency plans are
practicable;
(ii) These agencies are committed to
participating in any further
development of the plans, including any
required field demonstrations; and
(iii) These agencies are committed to
executing their responsibilities under
the plans in the event of an emergency.
(2) If certifications cannot be obtained
after sustained, good faith efforts by the
applicant, then the application must

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contain information, including a utility
plan, sufficient to show that the
proposed plans provide reasonable
assurance that adequate protective
measures can and will be taken in the
event of a radiological emergency at the
site.
(3) If complete and integrated
emergency plans were approved as part
of an early site permit, or submitted,
reviewed, and approved as part of the
CP application, new certifications that
demonstrate compliance with the
requirements of paragraph (i)(1) of this
section are not required.
(j) Organization. A description of the
applicant’s organizational structure,
allocations of responsibilities and
authorities, and personnel qualifications
requirements for operation.
(k) Maintenance program. A
description of a maintenance program
under § 53.715.
(l) Quality assurance. A description of
the QAP that demonstrates compliance
with the requirements under § 53.865.
(m) Radiation protection program. A
radiation protection program
description under § 53.850.
(n) Security program. A physical
security plan that describes how the
applicant will comply with § 53.860
(and 10 CFR part 11, if applicable,
including the identification and
description of jobs as required by
§ 11.11(a) of this chapter, at the
proposed facility). The plan must list
tests, inspections, audits, and other
means to be used to demonstrate
compliance with the requirements of 10
CFR parts 11 and 73, if applicable.
(o) Safeguards contingency plan. A
safeguards contingency plan in
accordance with the criteria set forth in
appendix C to 10 CFR part 73. The
safeguards contingency plan must
include plans for dealing with threats,
thefts, and radiological sabotage, as
defined in 10 CFR part 73, relating to
the SNM and nuclear facilities licensed
under this chapter and in the
applicant’s possession and control. Each
application for this type of license must
include the information contained in
the applicant’s safeguards contingency
plan. (Implementing procedures
required for this plan need not be
submitted for approval.) 1
(p) Security training and
qualification. A training and
qualification plan that describes how
the applicant will demonstrate
compliance with the criteria set forth in
§ 73.100 of this chapter or appendix B
to 10 CFR part 73.
(q) Cybersecurity plan. A
cybersecurity plan in accordance with
the criteria set forth in § 73.54 or
§ 73.110 of this chapter.

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(r) Security, safeguards and
cybersecurity plan implementation. A
description of the implementation of the
physical security plan, safeguards
contingency plan, training and
qualification plan, and cybersecurity
plan. Each applicant who prepares a
physical security plan, a safeguards
contingency plan, a training and
qualification plan, or a cybersecurity
plan must protect the plans and other
related Safeguards Information against
unauthorized disclosure in accordance
with the requirements of §§ 73.21 and
73.22 of this chapter.
(s) Fire protection program. A
description of the fire protection
program under § 53.875.
(t) Inservice inspection/inservice
testing program. A description of the
inservice inspection and inservice
testing programs under § 53.880.
(u) [Reserved]
(v) [Reserved]
(w) General employee training. A
description of the training program
required to demonstrate compliance
with § 53.830 and its implementation.
(x) Fitness-for-duty program. A
description of the FFD program required
by 10 CFR part 26 and its
implementation.
(y) Other programs. A description and
evaluation of the results of the
applicant’s programs, including
research and development, if any, to
demonstrate that any safety questions
identified at the CP stage have been
resolved.
(z) Safety design feature performance.
A description of how the performance of
each safety design feature has been
demonstrated capable of fulfilling
functional design criteria considering
interdependent effects through either
analysis, appropriate test programs,
prototype testing, operating experience,
or a combination thereof, in accordance
with § 53.440(a).
(aa) Technical specifications.
Proposed technical specifications
prepared in accordance with the
requirements of § 53.710(a).
1 A physical security plan that contains all
the information required in both § 73.55 or
§ 73.100 of this chapter and appendix C to 10
CFR part 73 satisfies the requirement for a
contingency plan.

§ 53.1372 Contents of applications for
operating licenses; other application
content.

In addition to the FSAR, the
application must also include the
following:
(a) Environmental report. An
environmental report in accordance
with § 51.53(b) of this chapter.
(b) Availability controls (if not
included in the FSAR). A description of

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the controls on plant operations,
including availability controls, to
provide reasonable confidence of safe
operation and that the configurations
and special treatments for SR and
NSRSS SSCs provide the capabilities
and reliabilities required to satisfy the
safety criteria of § 53.220, or more
restrictive alternative criteria adopted
under § 53.470, if not addressed by
Technical Specifications under
§ 53.1369(aa).
§ 53.1375

Review of applications.

(a) Standards for review of
applications. Applications filed under
this part will be reviewed according to
the standards set out in 10 CFR parts 20,
26, 51, 53, 73, and 140.
(b) Administrative review of
applications; hearings. A proceeding on
an OL is subject to all applicable
procedural requirements contained in
10 CFR part 2, including the
requirements for docketing (§ 2.101 of
this chapter) and issuance of a notice of
hearing (§ 2.104 of this chapter). All
hearings on OLs are governed by the
procedures contained in 10 CFR part 2.
§ 53.1381 Referral to the Advisory
Committee on Reactor Safeguards.

The Commission must refer a copy of
the application to the ACRS. The ACRS
must report on those portions of the
application that concern safety and
must apply the standards referenced in
§ 53.1375.

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§ 53.1384 Exemptions, departures, and
variances.

(a) Applicants for an OL under this
part, or any amendment to an OL, may
include in the application a request for
an exemption from one or more of the
Commission’s regulations. The
Commission may grant an exemption
request if it determines that the
exemption complies with § 53.080.
(b) An applicant for an OL who has
filed an application referencing an NRC
approval, permit, license, or
certification issued under this part may
include in the application a request for
departures, variances, or exemptions
related to the subject referenced NRC
approval, permit, license, or
certification. In determining whether to
grant the departure, variance, or
exemption, the Commission must apply
the same technically relevant criteria as
were applicable to the application for
the original or renewed approval,
license, or certification.
§ 53.1387

Issuance of operating licenses.

Upon completion of the construction
or alteration of a facility, in compliance
with the terms and conditions of the
construction permit and subject to any

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necessary testing of the facility for
health or safety purposes, the
Commission will, in the absence of good
cause shown to the contrary, issue an
OL or an appropriate amendment of the
license, as the case may be.
(a)(1) After receiving the report
submitted by the ACRS, the
Commission may issue an OL if the
Commission finds that—
(i) Construction of the facility has
been substantially completed in
conformity with the CP and the
application as amended, the provisions
of the Act, and the rules and regulations
of the Commission;
(ii) Any required notifications to other
agencies or bodies have been duly
made;
(iii) The facility will operate in
conformity with the application as
amended, the provisions of the Act, and
the rules and regulations of the
Commission;
(iv) There is reasonable assurance
that—
(A) the activities authorized by the OL
can be conducted without endangering
the health and safety of the public; and
(B) such activities will be conducted
in compliance with the regulations in
this chapter.
(v) The applicant is technically and
financially qualified to engage in the
activities authorized, however, no
finding of financial qualification is
necessary for an electric utility
applicant for an OL;
(vi) Issuance of the license will not be
inimical to the common defense and
security or to the health and safety of
the public;
(vii) The applicable provisions of 10
CFR part 140 have been satisfied; and
(viii) The findings required by subpart
A of 10 CFR part 51 have been made.
(2) [Reserved]
(b) [Reserved]
(c) The OL will include appropriate
provisions with respect to any
uncompleted items of construction and
such limitations or conditions as are
required to assure that operation during
the period of the completion of such
items will not endanger public health
and safety.
(d) The Commission will issue an OL
in such form and containing such
conditions and limitations, including
technical specifications, as it deems
necessary and appropriate.
§ 53.1390
licenses.

Backfitting of operating

After issuance of an OL, the
Commission may not modify, add, or
delete any term or condition of the OL,
except in accordance with the
provisions of § 53.1590.

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§ 53.1396

Duration of operating licenses.

The Commission will issue an OL
under this part for the term requested by
the applicant, not to exceed 40 years
from the date of issuance, or for the
estimated useful life of the facility if the
Commission determines that the
estimated useful life is less than the
term requested.
§ 53.1399

Transfer of an operating license.

An OL may be transferred under
§ 53.1570.
§ 53.1402

Application for renewal.

The filing of an application for a
renewed license must be in accordance
with § 53.1595.
§ 53.1405
license.

Continuation of an operating

Each OL for a facility that has
permanently ceased operations
continues in effect beyond the
expiration date to authorize ownership
and possession of the facility until the
Commission notifies the licensee in
writing that the license is terminated.
During this period of continued
effectiveness, the licensee must—
(a) Take actions necessary to
decommission and decontaminate the
facility and continue to maintain the
facility, including, where applicable, the
storage, control, and maintenance of the
spent fuel in a safe condition; and
(b) Conduct activities in accordance
with all other restrictions applicable to
the facility in accordance with the
NRC’s regulations and the provisions of
the OL for the facility.
§ 53.1410

Combined licenses.

Sections 53.1410 through 53.1461 set
out the requirements and procedures
applicable to Commission issuance of
COLs for commercial nuclear plants
under this part.
§ 53.1413 Contents of applications for
combined licenses; general information.

An application for a COL must
include the information required by
§ 53.1109 and the following
information:
(a) Except for an electric utility
applicant in regard to financial
assurance required after a Commission
finding under § 53.1452, the application
must include information sufficient to
demonstrate to the Commission the
financial qualification of the applicant
to carry out, in accordance with the
regulations in this chapter, the activities
for which the permit or license is
sought. As applicable, the following
should be provided:
(1) The applicant must submit
information that demonstrates that the
applicant possesses or has reasonable

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assurance of obtaining the funds
necessary to cover estimated
construction costs and related fuel cycle
costs. The applicant must submit
estimates of the total construction costs
of the facility and related fuel cycle
costs and must indicate the source(s) of
funds to cover these costs.
(2) The applicant must submit
information that demonstrates the
applicant possesses or has reasonable
assurance of obtaining the funds
necessary to cover estimated operation
costs for the period of the license. The
applicant must submit estimates for
total annual operating costs for each of
the first 5 years of operation of the
facility. The applicant must also
indicate the source(s) of funds to cover
these costs.
(3) Each application for a COL
submitted by a newly-formed entity
organized for the primary purpose of
constructing and operating a facility
must also include information
showing—
(i) The legal and financial
relationships the entity has or proposes
to have with its stockholders or owners;
and
(ii) The stockholders’ or owners’
financial ability to meet any contractual
obligation to the entity which they have
incurred or proposed to incur; and
(iii) Any other information considered
necessary by the Commission to enable
it to determine the applicant’s financial
qualification.
(4) The Commission may request an
established entity or newly formed
entity to submit additional or more
detailed information respecting its
financial arrangements and status of
funds if the Commission considers this
information appropriate. This may
include information regarding a
licensee’s ability to continue the
conduct of the activities authorized by
the license and to decommission the
facility.
(b) The application must include
information in the form of a report, as
described in subpart G of this part,
indicating how reasonable assurance
will be provided that funds will be
available to decommission the facility.

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§ 53.1416 Contents of applications for
combined licenses; technical information.

(a) Final Safety Analysis Report. The
application must contain an FSAR that
describes the facility and the limits on
its operation and presents a safety
analysis of the SSCs of the facility as a
whole. The Commission will require,
before issuance of a COL, that
engineering documents, such as
analyses, drawings, procurement
specifications, or construction and

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installation specifications, be completed
and available for audit if the more
detailed information is necessary for the
Commission to verify the information in
the application and make its safety
determination. The FSAR must include
the following information, at a level of
detail sufficient to enable the
Commission to reach a final conclusion
on all safety matters that must be
resolved by the Commission before
issuance of a COL:
(1) Site information. An application
for a COL for a commercial nuclear
reactor must include the site
information required for an early site
permit in § 53.1146(a)(1)(iv) through (x).
(2) Design information. An
application for a COL for a commercial
nuclear plant must include the design
information equivalent to that required
for a standard design certification as
defined in § 53.1239(a)(2) through (7),
(a)(9), and (a)(11) through (27).
(3) Technical qualifications. A
description of the technical
qualifications of the applicant to engage
in the proposed activities in accordance
with the regulations in this chapter.
(4) Integrity assessment program. A
description of an Integrity Assessment
Program that addresses the elements
described in § 53.870.
(5) Safeguards information. A
description of the program to protect
Safeguards Information against
unauthorized disclosure in accordance
with the requirements in §§ 73.21 and
73.22 of this chapter, as applicable.
(6) Emergency response facility or
facilities. Description of the locations
and capabilities to be established for
command and control, support, and
coordination of onsite and offsite, as
applicable, functions during reactor
accident conditions.
(7) Role of personnel. (i) A description
of the completed assessments related to
the role of personnel in ensuring safe
operations considering the analyses
required by § 53.730. These assessments
must include the following:
(A) Human factors engineering design
requirements of § 53.730(a);
(B) Human system interface design
requirements of § 53.730(b);
(C) Concept of operations of
§ 53.730(c); and
(D) Functional requirements analysis
and function allocation of § 53.730(d);
(ii) A description of the program to be
used for evaluating and applying
operating experience as required by
§ 53.730(e);
(iii) A staffing plan and supporting
analyses as required by § 53.730(f).
(8) Training, examination, and
proficiency programs. (i) A description
of the training, examination, and

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87097

proficiency programs required by
§ 53.730(g); and
(ii) A description of the training
programs required by § 53.830.
(9) Emergency plan. Emergency plans
complying with the requirements of
§ 53.855.
(i) The emergency plan must include,
as applicable, all emergency plan
certifications that have been obtained
from the State, local, and participating
Tribal governmental agencies with
emergency planning responsibilities.
The certifications must state that—
(A) The proposed emergency plans
are practicable;
(B) These agencies are committed to
participating in any further
development of the plans, including any
required field demonstrations; and
(C) These agencies are committed to
executing their responsibilities under
the plans in the event of an emergency.
(ii) If certifications cannot be obtained
after sustained, good faith efforts by the
applicant, then the application must
contain information, including a utility
plan, sufficient to show that the
proposed plans provide reasonable
assurance that adequate protective
measures can and will be taken in the
event of a radiological emergency at the
site.
(10) Organization. A description of
the applicant’s organizational structure,
allocations of responsibilities and
authorities, and personnel qualifications
requirements for operation.
(11) Maintenance program. A
description of a maintenance program
under § 53.715.
(12) Quality assurance. A description
of the QAP under § 53.865.
(13) Radiation protection program. A
radiation protection program
description under § 53.850.
(14) Security program. A physical
security plan that describes how the
applicant will comply with § 53.860
(and 10 CFR part 11, if applicable,
including the identification and
description of jobs as required by
§ 11.11(a) of this chapter, at the
proposed facility). The plan must list
tests, inspections, audits, and other
means to be used to demonstrate
compliance with the requirements of 10
CFR parts 11 and 73, if applicable.
(15) Safeguards contingency plan. A
safeguards contingency plan in
accordance with the criteria set forth in
appendix C to 10 CFR part 73. The
safeguards contingency plan must
include plans for dealing with threats,
thefts, and radiological sabotage, as
defined in 10 CFR part 73, relating to
the SNM and nuclear facilities licensed
under this chapter and in the
applicant’s possession and control. Each

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application for this type of license must
include the information contained in
the applicant’s safeguards contingency
plan.1 (Implementing procedures
required for this plan need not be
submitted for approval.)
(16) Security training and
qualification. A training and
qualification plan that describes how
the applicant will demonstrate
compliance with the criteria set forth in
§ 73.100 of this chapter or appendix B
to 10 CFR part 73.
(17) Cybersecurity plan. A
cybersecurity plan in accordance with
the criteria set forth in § 73.54 or
§ 73.110 of this chapter.
(18) Security, safeguards and
cybersecurity plan implementation. A
description of the implementation of the
physical security plan, safeguards
contingency plan, training and
qualification plan, and cybersecurity
plan. Each applicant who prepares a
physical security plan, a safeguards
contingency plan, a training and
qualification plan, or a cybersecurity
plan must protect the plans and other
related Safeguards Information against
unauthorized disclosure in accordance
with the requirements of §§ 73.21 and
73.22 of this chapter.
(19) Fire protection program. A
description of the fire protection
program under § 53.875.
(20) Inservice inspection/inservice
testing program. Descriptions of
inservice inspection and inservice
testing programs under § 53.880.
(21) [Reserved]
(22) [Reserved]
(23) General employee training. A
description of the training program
required to demonstrate compliance
with § 53.830 and its implementation.
(24) Fitness-for-duty program. A
description of the FFD program under
part 26 of this chapter and its
implementation.
(25) Technical specifications.
Proposed technical specifications
prepared in accordance with the
requirements of § 53.710(a).
(b) If there are SSCs of the plant for
which research and development is
necessary to confirm the adequacy of
their design, a report which documents
the resolution of any safety questions
associated with such SSCs.
(c) A description of how the
performance of each safety design
feature has been demonstrated capable
of fulfilling functional design criteria
considering interdependent effects
through either analysis, appropriate test
programs, prototype testing, operating
experience, or a combination thereof, in
accordance with § 53.440(a).

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(d) If the COL application references
an early site permit, then the following
requirements apply:
(1) The FSAR need not contain
information or analyses submitted to the
Commission in connection with the
early site permit provided that the FSAR
must either include or incorporate by
reference the early site permit Site
Safety Analysis Report and contain, in
addition to the information and analyses
otherwise required, information
sufficient to demonstrate that the design
of the facility falls within the site
characteristics and design parameters
specified in the early site permit.
(2) If the FSAR does not demonstrate
that design of the facility falls within
the site characteristics and design
parameters, the application must
include a request for a variance that
complies with the requirements of
§§ 53.1188(d) and 53.1437.
(3) The FSAR must demonstrate that
all terms and conditions that have been
included in the early site permit will be
satisfied by the date of issuance of the
COL. Any terms or conditions of the
early site permit that could not be met
by the time of issuance of the COL must
be set forth as terms or conditions of the
COL.
(4) If the early site permit approves
complete and integrated emergency
plans, or major features of emergency
plans, then the FSAR must include any
new or additional information that
updates and corrects the information
that was provided under § 53.1146(b)(2)
and discuss whether the new or
additional information materially
changes the bases for compliance with
the applicable requirements. The
application must identify changes to the
emergency plans or major features of
emergency plans that have been
incorporated into the proposed facility
emergency plans and that constitute or
would constitute a change in an
emergency plan that results in reducing
the licensee’s capability to perform an
emergency planning function in the
event of a radiological emergency.
(5) If complete and integrated
emergency plans are approved as part of
the early site permit, new certifications
meeting the requirements of paragraph
(a)(9)(i) of this section are not required.
(e) If the COL application references
a standard design approval, then the
following requirements apply:
(1) The FSAR need not contain
information or analyses submitted to the
Commission in connection with the
design approval, provided, however,
that the FSAR must either include or
incorporate by reference the standard
design approval FSAR and must
contain, in addition to the information

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and analyses otherwise required,
information sufficient to demonstrate
that the characteristics of the site fall
within the site parameters specified in
the design approval. In addition, the
plant-specific PRA information must
use the PRA information for the design
approval and must be updated to
account for site specific design
information and any design changes or
departures.
(2) The FSAR must demonstrate that
all terms and conditions that have been
included in the design approval will be
satisfied by the date of issuance of the
COL.
(f) If the COL application references a
standard design certification, then the
following requirements apply:
(1) The FSAR need not contain
information or analyses submitted to the
Commission in connection with the
standard design certification, provided,
however, that the FSAR must either
include or incorporate by reference the
standard design certification FSAR and
must contain, in addition to the
information and analyses otherwise
required, information sufficient to
demonstrate that the site characteristics
fall within the site parameters specified
in the standard design certification. In
addition, the plant-specific PRA
information must use the PRA
information for the standard design
certification and must be updated to
account for site-specific design
information and any design changes or
departures.
(2) The FSAR must demonstrate that
the interface requirements established
for the design under § 53.1239(a)(24)
have been met.
(3) The FSAR must demonstrate that
all requirements and restrictions set
forth in the referenced standard design
certification rule must be satisfied by
the date of issuance of the COL. Any
requirements and restrictions set forth
in the referenced standard design
certification rule that could not be
satisfied by the time of issuance of the
COL, must be set forth as terms or
conditions of the COL.
(g) If the COL application references
the use of one or more manufactured
reactors licensed under § 53.1270, then
the following requirements apply:
(1) The FSAR need not contain
information or analyses submitted to the
Commission in connection with the ML,
provided, however, that the FSAR must
either include or incorporate by
reference the ML FSAR and must
contain, in addition to the information
and analyses otherwise required,
information sufficient to demonstrate
that the site characteristics fall within
the site parameters specified in the ML.

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In addition, the plant-specific PRA
information must use the PRA
information for the manufactured
reactor and must be updated to account
for site-specific design information and
any design changes or departures.
(2) The FSAR must demonstrate that
the interface requirements established
for the design have been met.
(3) The FSAR must demonstrate that
all terms and conditions that have been
included in the ML will be satisfied by
the date of issuance of the COL. Any
terms or conditions of the ML that could
not be met by the time of issuance of the
COL, must be set forth as terms or
conditions of the COL.
(h) Each applicant for a COL under
this part must protect Safeguards
Information against unauthorized
disclosure in accordance with the
requirements in §§ 73.21 and 73.22 of
this chapter, as applicable.
1 A physical security plan that contains all
the information required in both § 73.55 or
§ 73.100 of this chapter and appendix C to 10
CFR part 73 demonstrates compliance with
the requirement for a contingency plan.

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§ 53.1419 Contents of applications for
combined licenses; other application
content.

(a) In addition to the FSAR, the
application must also include the
following:
(1) Environmental report. (i) An
environmental report either in
accordance with § 51.50(c) of this
chapter if an LWA under § 53.1130 is
not requested in conjunction with the
COL application, or in accordance with
§§ 51.49 and 51.50(c) of this chapter if
an LWA is requested in conjunction
with the COL application; or
(ii) If the applicant wishes to request
that an LWA under § 53.1130 be issued
before issuance of the COL, the
information otherwise required by
§ 53.1130, in accordance with either
§ 2.101(a)(1) through (a)(4), or
§ 2.101(a)(9) of this chapter;
(2) Availability controls (if not
included in the FSAR). A description of
the controls on plant operations,
including availability controls, to
provide reasonable confidence of safe
operation and that the configurations
and special treatments for SR SSCs and
NSRSS SSCs provide the capabilities
and reliabilities required to satisfy the
safety criteria of § 53.220, or more
restrictive alternative criteria adopted
under § 53.470, if not addressed by
Technical Specifications under
§ 53.1416(a)(25); and
(3) Inspections, tests, analyses, and
acceptance criteria. The proposed
inspections, tests, and analyses,
including those applicable to emergency

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planning, that the licensee must
perform, and the acceptance criteria that
are necessary and sufficient to provide
reasonable assurance that, if the
inspections, tests, and analyses are
performed and the acceptance criteria
met, the facility has been constructed
and will be operated in conformity with
the COL, the provisions of the Act, and
the Commission’s rules and regulations.
(i) If the application references an
early site permit with ITAAC, the early
site permit ITAAC must apply to those
aspects of the COL which are approved
in the early site permit.
(ii) If the application references a
standard design certification, the ITAAC
contained in the certified design must
apply to those portions of the facility
design which are approved in the
standard design certification.
(iii) If the application references an
ML, the ITAAC contained in the ML
must apply to those portions of the
facility design which are approved in
the ML.
(iv) If the application references an
early site permit with ITAAC, a
standard design certification, an ML, or
combination thereof, the application
may include a notification that a
required inspection, test, or analysis in
the ITAAC has been successfully
completed and that the corresponding
acceptance criterion has been met. The
Federal Register notification required
by § 53.1422 of this chapter must
indicate that the application includes
this notification.
(b) [Reserved]
§ 53.1422

Review of applications.

(a) Standards for review of
applications. Applications filed under
this part will be reviewed according to
the standards set out in this part and 10
CFR parts 20, 51, 73, and 140.
(b) Administrative review of
applications; hearings. A proceeding on
a COL is subject to all applicable
procedural requirements contained in
10 CFR part 2, including the
requirements for docketing (§ 2.101 of
this chapter) and issuance of a notice of
hearing (§ 2.104 of this chapter). If an
applicant requests a Commission
finding on certain ITAAC with the
issuance of the COL, then those ITAAC
will be identified in the notice of
hearing. All hearings on COLs are
governed by the procedures contained
in 10 CFR part 2.
§ 53.1425 Finality of referenced NRC
approvals.

If the application for a COL under this
part references an early site permit,
standard design certification rule,
standard design approval, or ML, issued

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under this part, the scope and nature of
matters resolved for the application and
any COL issued are governed by the
relevant provisions addressing finality,
including §§ 53.1188, 53.1221, 53.1263,
and 53.1288.
§ 53.1431 Referral to the Advisory
Committee on Reactor Safeguards.

The Commission must refer a copy of
the application to the ACRS. The ACRS
must report on those portions of the
application that concern safety and
must apply the standards referenced in
§ 53.1422, in accordance with the
finality provisions in § 53.1425.
§ 53.1434 Authorization to conduct limited
work authorization activities.

(a) If the application for a COL under
this part does not reference an early site
permit which authorizes the holder to
perform the activities under
§ 53.1130(b), the applicant may not
perform those activities without
obtaining the separate authorization
required by § 53.1130(a). Authorization
may be granted only after the presiding
officer in the proceeding on the
application has made the findings and
determination required by
§ 53.1130(b)(1)(ii) and (b)(1)(iv), and the
Director, Office of Nuclear Reactor
Regulation makes the determination
required by § 53.1130(b)(1)(iii).
(b) If, after an applicant has performed
the activities permitted by a LWA
issued under § 53.1130, the application
for the COL is withdrawn or denied,
then the applicant must implement the
approved site redress plan.
§ 53.1437 Exemptions, departures, and
variances.

(a) An applicant for a COL, or any
amendment to a COL, may include in
the application a request for an
exemption from one or more of the
Commission’s regulations.
(1) If the request is for an exemption
from any part of a referenced standard
design certification rule, the
Commission may grant the request if it
determines that the exemption complies
with any exemption provisions of the
referenced standard design certification
rule, or with § 53.1263 if there are no
applicable exemption provisions in the
referenced standard design certification
rule.
(2) For all other requests for
exemptions, the Commission may grant
a request if it determines that the
exemption complies with § 53.080.
(b) An applicant for a COL who has
filed an application referencing an early
site permit issued under § 53.1158 may
include in the application a request for
a variance from one or more site
characteristics, design parameters, or

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terms and conditions of the permit, or
from the Site Safety Analysis Report. In
determining whether to grant the
variance, the Commission must apply
the same technically relevant criteria as
were applicable to the application for
the original or renewed site permit.
Once a COL referencing an early site
permit is issued, variances from the
early site permit will not be granted for
that CP or COL.
(c) An applicant for a COL who has
filed an application referencing use of a
manufactured reactor may include in
the application a request for a departure
from one or more design characteristics,
site parameters, terms and conditions,
or approved design of the manufactured
reactor under the ML issued under
§ 53.1287. The Commission may grant
such a request only if it determines that
the departure will comply with the
requirements of § 53.080, and that the
special circumstances outweigh any
decrease in safety that may result from
the reduction in standardization caused
by the departure.
(d) Issuance of a variance under
paragraph (b) of this section or a
departure under paragraph (c) of this
section is subject to litigation during the
COL proceeding in the same manner as
other issues material to that proceeding.

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§ 53.1440

Issuance of combined licenses.

(a)(1) After conducting a hearing
under § 53.1422(b) and receiving the
report submitted by the ACRS, the
Commission may issue a COL if the
Commission finds that—
(i) The applicable standards and
requirements of the Act and the
Commission’s regulations have been
met;
(ii) Any required notifications to other
agencies or bodies have been duly
made;
(iii) There is reasonable assurance that
the facility will be constructed and will
operate in conformity with the license,
the provisions of the Act, and the
Commission’s regulations;
(iv) The applicant is technically and
financially qualified to engage in the
activities authorized; however, no
finding of financial qualification is
necessary for an electric utility
applicant for a COL;
(v) Issuance of the license will not be
inimical to the common defense and
security or to the health and safety of
the public; and
(vi) The findings required by subpart
A of 10 CFR part 51 have been made.
(2) The Commission may also find, at
the time it issues the COL, that certain
acceptance criteria in one or more of the
ITAAC in a referenced early site permit,
standard design certification, or ML

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have been met. This finding will finally
resolve that those acceptance criteria
have been met, those acceptance criteria
will be deemed to be excluded from the
COL, and findings under § 53.1452(g)
with respect to those acceptance criteria
are unnecessary.
(b) The Commission must identify
within the COL the inspections, tests,
and analyses, including those applicable
to emergency planning, that the licensee
must perform, and the acceptance
criteria that, if met, are necessary and
sufficient to provide reasonable
assurance that the facility has been
constructed and will be operated in
conformity with the license, the
provisions of the Act, and the
Commission’s rules and regulations.
(c) A COL must contain the terms and
conditions, including technical
specifications, as the Commission
deems necessary and appropriate.
§ 53.1443

Finality of combined licenses.

(a) After issuance of a COL, the
Commission may not modify, add, or
delete any term or condition of the COL,
the design of the facility, the ITAAC
contained in the license that are not
derived from a referenced standard
design certification or ML, except under
the provisions of § 53.1452 or § 53.1590.
(b) If the COL does not reference a
standard design certification or use of a
manufactured reactor under an ML
issued under § 53.1287, then a licensee
may make changes in the facility as
described in the FSAR (as updated) and
make changes in the procedures as
described in the FSAR (as updated)
under the applicable change processes
in § 53.1550.
(c) If the COL references a certified
design, then—
(1) Changes to or departures from
information within the scope of the
referenced standard design certification
rule are subject to the applicable change
processes in that rule; and
(2) Changes that are not within the
scope of the referenced standard design
certification rule are subject to the
applicable change processes in subpart
I of this part, unless they also involve
changes to or noncompliance with
information within the scope of the
referenced standard design certification
rule. In these cases, the applicable
provisions of this section and the
standard design certification rule apply.
(d) If the COL references use of a
manufactured reactor under an ML
issued under this part, then—
(1) Changes to or departures from
information within the scope of the
manufactured reactor’s design are
subject to the change processes in
§ 53.1288; and

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(2) Changes that are not within the
scope of the manufactured reactor’s
design are subject to the applicable
change processes in subpart I.
(e) The Commission may issue and
make immediately effective any
amendment to a COL upon a
determination by the Commission that
the amendment involves no significant
hazards consideration, notwithstanding
the pendency before the Commission of
a request for a hearing from any person.
The amendment may be issued and
made immediately effective in advance
of the holding and completion of any
required hearing. The amendment will
be processed under the procedures
specified in § 53.1515.
(f) Any modification to, addition to, or
deletion from the terms and conditions
of a COL, including any modification to,
addition to, or deletion from the
inspections, tests, and analyses, or
related acceptance criteria contained in
the license is a proposed amendment to
the license. There must be an
opportunity for a hearing on the
amendment.
§ 53.1449

Inspection during construction.

(a) Licensee schedule for inspections,
tests, or analyses. The licensee must
submit to the NRC, no later than 1 year
after issuance of the COL or at the start
of construction as defined at § 53.020,
whichever is later, its schedule for
completing the inspections, tests, or
analyses in the ITAAC. The licensee
must submit updates to the ITAAC
schedules every 6 months thereafter
and, within 1 year of its scheduled date
for initial loading of fuel (or, for a fueled
manufactured reactor, within 1 year of
its scheduled date for initiating the
physical removal of any one of the
independent physical mechanisms to
prevent criticality required under
§ 53.620(d)(1)), the licensee must submit
updates to the ITAAC schedule every 30
days until the final notification is
provided to the NRC under paragraph
(c)(1) of this section.
(b) Licensee and applicant conduct of
activities subject to ITAAC. With respect
to activities subject to an ITAAC, an
applicant for a COL may proceed at its
own risk with design and procurement
activities, and a licensee may proceed at
its own risk with design, procurement,
construction, and preoperational
activities, even though the NRC may not
have found that any one of the
prescribed acceptance criteria are met.
(c) Licensee notifications. (1) ITAAC
closure notification. The licensee must
notify the NRC that prescribed
inspections, tests, and analyses have
been performed and that the prescribed
acceptance criteria are met. The

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notification must contain sufficient
information to demonstrate that the
prescribed inspections, test, and
analyses have been performed and that
the prescribed acceptance criteria are
met.
(2) ITAAC post-closure notifications.
Following the licensee’s ITAAC closure
notifications under paragraph (c)(1) of
this section until the Commission makes
the finding under § 53.1452(g), the
licensee must notify the NRC, in a
timely manner, of new information that
materially alters the basis for
determining that either inspections,
tests, or analyses were performed as
required, or that acceptance criteria are
met. The notification must contain
sufficient information to demonstrate
that, notwithstanding the new
information, the prescribed inspections,
tests, and analyses have been performed
as required, and the prescribed
acceptance criteria are met.
(3) Uncompleted ITAAC notification.
If the licensee has not provided, by the
date 225 days before the scheduled date
for initial loading of fuel (or, for a fueled
manufactured reactor, by the date 225
days before the scheduled date for
initiating the physical removal of any
one of the independent physical
mechanisms to prevent criticality
required under § 53.620(d)(1)), the
notification required by paragraph (c)(1)
of this section for all ITAAC, then the
licensee must notify the NRC that the
prescribed inspections, tests, or analyses
for all uncompleted ITAAC will be
performed and that the prescribed
acceptance criteria will be met prior to
operation. The notification must be
provided no later than the date 225 days
before the scheduled date for initial
loading of fuel (or, for a fueled
manufactured reactor, no later than the
date 225 days before the scheduled date
for initiating the physical removal of
any one of the independent physical
mechanisms to prevent criticality
required under § 53.620(d)(1)), and must
provide sufficient information to
demonstrate that the prescribed
inspections, tests, or analyses will be
performed and the prescribed
acceptance criteria for the uncompleted
ITAAC will be met, including, but not
limited to, a description of the specific
procedures and analytical methods to be
used for performing the prescribed
inspections, tests, and analyses and
determining that the prescribed
acceptance criteria are met.
(4) All ITAAC complete notification.
The licensee must notify the NRC that
all ITAAC are complete.
(d) Licensee determination of
noncompliance with ITAAC. (1) In the
event that an activity is subject to an

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ITAAC derived from a referenced
standard design certification and the
licensee has not demonstrated that the
prescribed acceptance criteria are met,
the licensee may take corrective actions
to successfully complete that ITAAC or
request an exemption from the standard
design certification ITAAC, as
applicable. A request for an exemption
must also be accompanied by a request
for a license amendment under subpart
I.
(2) In the event that an activity is
subject to an ITAAC not derived from a
referenced standard design certification
and the licensee has not demonstrated
that the prescribed acceptance criteria
are met, the licensee may take corrective
actions to successfully complete that
ITAAC or request a license amendment
under subpart I.
(e) NRC inspection, publication of
notices, and availability of licensee
notifications. The NRC must ensure that
the prescribed inspections, tests, and
analyses in the ITAAC are performed.
(1) At appropriate intervals until the
last date for submission of requests for
hearing under § 53.1452, the NRC must
publish notices in the Federal Register
of the NRC staff’s determination of the
successful completion of inspections,
tests, and analyses.
(2) The NRC must make publicly
available the licensee notifications
under paragraph (c) of this section. The
NRC must, no later than the date of
publication of the notice of intended
operation required by § 53.1452(a),
make publicly available those licensee
notifications under paragraph (c) of this
section that have been submitted to the
NRC at least 7 days before that notice.
§ 53.1452
license.

Operation under a combined

(a) The licensee must notify the NRC
of its scheduled date for initial loading
of fuel no later than 270 days before the
scheduled date and must notify the NRC
of updates to its schedule every 30 days
thereafter.1 Not less than 180 days
before the date scheduled for initial
loading of fuel into a plant by a licensee
that has been issued a COL under this
part, the Commission must publish
notice of intended operation in the
Federal Register.2 The notice must
provide that any person whose interest
may be affected by operation of the
plant may, within 60 days, request that
the Commission hold a hearing on
whether the facility as constructed
complies, or on completion will
comply, with the acceptance criteria in
the COL, except that a hearing must not
be granted for those ITAAC that the
Commission found were met under
§ 53.1440(a)(2).

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(b) A request for hearing under
paragraph (a) of this section must show,
prima facie—
(1) That one or more of the acceptance
criteria of the ITAAC in the COL have
not been, or will not be, met; and
(2) The specific operational
consequences of nonconformance that
would be contrary to providing
reasonable assurance of adequate
protection of the public health and
safety.
(c) The Commission, acting as the
presiding officer, must determine
whether to grant or deny the request for
hearing under the applicable
requirements of § 2.309 of this chapter.
If the Commission grants the request,
the Commission, acting as the presiding
officer, must determine whether during
a period of interim operation there will
be reasonable assurance of adequate
protection to the public health and
safety. The Commission’s determination
must consider the petitioner’s prima
facie showing and any answers thereto.
If the Commission determines there is
such reasonable assurance, it must
allow operation during an interim
period under the COL.
(d) The Commission, in its discretion,
must determine appropriate hearing
procedures, whether informal or formal
adjudicatory, for any hearing under
paragraph (a) of this section, and must
state its reasons therefore.
(e) The Commission must, to the
maximum possible extent, render a
decision on issues raised by the hearing
request within 180 days of the
publication of the notice provided by
paragraph (a) of this section or by the
anticipated date for initial loading of
fuel into the reactor, whichever is later.
(f) A petition to modify the terms and
conditions of the COL will be processed
as a request for action under § 2.206 of
this chapter. The petitioner must file the
petition with the Secretary of the
Commission. Before the licensed
activity allegedly affected by the
petition (fuel loading, low power
testing, etc.) commences, the
Commission must determine whether
any immediate action is required. If the
petition is granted, then an appropriate
order will be issued. Fuel loading and
operation under the COL will not be
affected by the granting of the petition
unless the order is made immediately
effective.
(g) The licensee must not operate the
facility until the Commission makes a
finding that the acceptance criteria in
the COL are met, except for those
acceptance criteria that the Commission
found were met under § 53.1440(a)(2). If
the COL is for a modular design, each

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reactor unit may require a separate
finding as construction proceeds.
(h) After the Commission has made
the finding in paragraph (g) of this
section, the ITAAC do not, by virtue of
their inclusion in the COL, constitute
regulatory requirements either for
licensees or for renewal of the license;
except for the specific ITAAC for which
the Commission has granted a hearing
under paragraph (a) of this section, all
ITAAC expire upon final Commission
action in the proceeding. However,
subsequent changes to the facility or
procedures described in the FSAR (as
updated) must comply with the
requirements in § 53.1443(e) or (f), as
applicable.
1 For licensees installing fueled
manufactured reactors under a COL, the COL
holder must instead notify the NRC of its
scheduled date for initiating the physical
removal of any one of the independent
physical mechanisms to prevent criticality
required under § 53.620(d)(1) no later than
270 days before the scheduled date and must
notify the NRC of updates to its schedule
every 30 days thereafter.
2 For licensees installing fueled
manufactured reactors under a COL, the
Commission must instead publish notice of
intended operation in the Federal Register
not less than 180 days before the date
scheduled for initiating the physical removal
of any one of the independent physical
mechanisms to prevent criticality required
under § 53.620(d)(1).

§ 53.1455

Duration of combined license.

A COL is issued for a specified period
not to exceed 40 years from the date on
which the Commission makes a finding
that acceptance criteria are met under
§ 53.1452(g) or allowing operation
during an interim period under the COL
under § 53.1452(c).
§ 53.1456

Transfer of a combined license.

A COL may be transferred under
§ 53.1570.
§ 53.1458

Application for renewal.

The filing of an application for a
renewed license must be in accordance
with § 53.1595.

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§ 53.1461
license.

Continuation of combined

Each COL for a facility that has
permanently ceased operations
continues in effect beyond the
expiration date to authorize ownership
and possession of the facility until the
Commission notifies the licensee in
writing that the license is terminated.
During this period of continued
effectiveness, the licensee must—
(a) Take actions necessary to
decommission and decontaminate the
facility and continue to maintain the
facility, including, where applicable, the

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storage, control and maintenance of the
spent fuel, in a safe condition; and
(b) Conduct activities in accordance
with all other restrictions applicable to
the facility in accordance with the
NRC’s regulations and the provisions of
the COL for the facility.
§ 53.1470 Standardization of commercial
nuclear plant designs: licenses to construct
and operate nuclear power reactors of
identical design at multiple sites.

(a) Except as otherwise specified in
this section, the provisions of this
section apply to CP, OL, and COL
applications for commercial nuclear
plants of identical design (the ‘‘common
design’’) under this part.
(b) Each application for a CP, OL, or
COL submitted pursuant to this section
must be submitted as specified in
§§ 53.1300, 53.1360, or 53.1410,
respectively, and § 2.101 of this chapter.
Each application must state that the
applicant wishes to construct a facility
identical to a facility proposed for one
or more sites other than the applicant’s
(the ‘‘common design’’), and the
applicant wishes to have the application
considered under this section. Each
application must list each of the other
applications to be treated together under
this section.
(c) Each application must include the
information required by the applicable
sections of this subpart, provided
however, that the application must
identify the common design, and, if
applicable, reference a standard design
certification or standard design approval
under this part, or the use of a reactor
manufactured under this part. The
FSAR for each application must either
incorporate by reference or include the
final safety analysis of the common
design, including, if applicable, the
FSAR for the referenced standard design
certification, standard design approval,
or the manufactured reactor.
(d) Each application submitted
pursuant to this section must contain an
environmental report under
§§ 53.1312(a)(1), 53.1372(a), or
53.1419(a)(1), as applicable, that
complies with the applicable provisions
of 10 CFR part 51, provided, however,
that the application may incorporate by
reference a single environmental report
on the environmental impacts of the
common design that are applicable to
each site.
(e) Upon a determination that each
application is acceptable for docketing
under § 2.101 of this chapter, each
application will be docketed and a
notice of docketing for each application
will be published in the Federal
Register, under § 2.104 of this chapter,
provided, however, that the notice must

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state that the application will be
processed under the provisions of this
section and subpart D of 10 CFR part 2.
At the discretion of the Commission, a
single notice of docketing for multiple
applications may be published in the
Federal Register.
(f) The NRC must prepare an
environmental assessment or draft and
final environmental impact statements
for each of the applications under 10
CFR part 51. Scoping under §§ 51.28
and 51.29 of this chapter for each of the
license applications may be conducted
simultaneously and joint scoping may
be conducted with respect to the
environmental issues relevant to the
common design. If the applications
reference a standard design certification,
then the environmental assessment or
environmental impact statement for
each of the applications must
incorporate by reference the standard
design certification environmental
assessment. If the applications do not
reference a standard design certification,
then the NRC must prepare
environmental assessments or draft and
final supplemental environmental
impact statements which address severe
accident mitigation design alternatives
for the common design, which must be
incorporated by reference into the
environmental assessment or
environmental impact statement
prepared for each application. Scoping
under §§ 51.28 and 51.29 of this chapter
for the supplemental environmental
impact statement may be conducted
simultaneously and may be part of the
scoping for each of the applications.
(g) The ACRS must report on each of
the applications as required by the
applicable sections of this subpart. Each
report must be limited to those safety
matters for each application that are not
relevant to the common design. In
addition, the ACRS must separately
report on the safety of the common
design, provided, however, that the
report need not address the safety of a
referenced standard design certification
or reactor manufactured under this part.
(h) The Commission must designate a
presiding officer to conduct the
proceeding with respect to the health
and safety, common defense and
security, and environmental matters
relating to the common design and
affecting at least two applications. The
hearing will be governed by the
applicable provisions of subparts A, C,
G, L, N, and O of 10 CFR part 2 relating
to applications for CPs, OLs, and COLs.
The presiding officer must issue a
partial initial decision on the common
design.
(i) If the design for the power
reactor(s) proposed in a particular

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application is not identical to the others,
that application may not be processed
under this section and subpart D of 10
CFR part 2.
(j) As used in this section, the design
of a nuclear power reactor included in
a single referenced Safety Analysis
Report means the design of those SSCs
important to radiological health and
safety and the common defense and
security.
Subpart I—Maintaining and Revising
Licensing-Basis Information
§ 53.1500

Licensing-basis information.

This subpart provides the
requirements for each holder of a
license for a commercial nuclear plant
licensed under this part to maintain
licensing-basis information as defined
in § 53.020; evaluate changes to site
characteristics, plant design features,
and programmatic controls to determine
needed approvals and revisions; and
submit appropriate updates to the U.S.
Nuclear Regulatory Commission (NRC).

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§ 53.1502
licenses.

Specific terms and conditions of

(a) Each license issued under this part
is subject to the provisions of the
Atomic Energy Act of 1954, as amended,
(the Act) and to all rules, regulations,
and orders of the Commission. The
terms and conditions of the license will
be subject to amendment, revision, or
modification, by reason of amendments
of the Act or by reason of rules,
regulations, and orders issued in
accordance with the terms of the Act.
(b) Each license issued under this part
must be subject to all conditions
imposed as a matter of law by sections
401(a)(2) and 401(d) of the Federal
Water Pollution Control Act, as
amended (33 U.S.C.A. 1341(a)(2) and
(d)).
(c) A holder of an operating license
(OL) or combined license (COL) under
this part may take reasonable action that
departs from a license condition or a
technical specification included in a
license issued under this part in a
national security emergency established
by a law enacted by the Congress or by
an order or directive issued by the
President pursuant to statutes or the
Constitution of the United States. The
authority under this paragraph must be
exercised in accordance with law,
including section 57e of the Act, and is
in addition to the authority granted
under § 53.740(h), which remains in
effect unless otherwise directed by the
Commission during a national security
emergency. The authority under this
paragraph may be exercised—
(1) When this action is immediately
needed to implement national security

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objectives as designated by the national
command authority through the
Commission; and
(2) No action consistent with license
conditions and technical specifications
that can satisfy national security
objectives is immediately apparent.
(d)(1) If the NRC finds that the state
of emergency preparedness does not
provide reasonable assurance that
adequate protective measures can and
will be taken in the event of a
radiological emergency (including
findings based on requirements of 10
CFR part 50, appendix E, section IV.D.3)
and if the deficiencies (including
deficiencies based on requirements of
10 CFR part 50, appendix E, section
IV.D.3) are not corrected within 4
months of that finding, the Commission
will determine whether the facility must
be shut down or cease operations until
such deficiencies are remedied or
whether other enforcement action is
appropriate. In determining whether a
shutdown or other enforcement action is
appropriate, the Commission will take
into account, among other factors,
whether the licensee can demonstrate to
the Commission’s satisfaction that the
deficiencies in the plan are not
significant for the plant in question, or
that adequate interim compensating
actions have been or will be taken
promptly, or that that there are other
compelling reasons for continued
operation.
(2) If the planning standards for
radiological emergency preparedness
apply to offsite emergency response
plans, or if the planning activities in
§ 50.160(b)(1)(iv)(B) apply, then the
NRC will base its finding on a review of
the Federal Emergency Management
Agency findings and determinations as
to whether State, participating Tribal
and local emergency plans are adequate
and capable of being implemented, and
on the NRC assessment as to whether
the licensee’s emergency plans are
adequate and capable of being
implemented. Nothing in this paragraph
must be construed as limiting the
authority of the Commission to take
action under any other regulation or
authority of the Commission or at any
time other than that specified in this
paragraph.
§ 53.1505 Changes to licensing-basis
information requiring prior NRC approval.

(a) Sections 53.1510 through 53.1520
provide the process for a licensee to
request and the NRC to issue
amendments to licenses, including any
conditions contained therein, technical
specifications or other attachments to a
license, and any orders issued by the
NRC modifying a license. Sections

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53.1525 and 53.1530 govern proposed
changes to a commercial nuclear plant
referencing a certified design or
manufacturing license (ML).
(b) A licensee may propose changing
licensing-basis information established
by NRC regulations by requesting an
exemption in accordance with § 53.080.
§ 53.1510
license.

Application for amendment of

Whenever a holder of a license under
this part desires to amend the license,
an application for an amendment must
be filed with the Commission, as
specified in § 53.040, that fully
describes the changes desired and,
following as far as applicable, the form
prescribed for original applications.
Applications for amendments involving
changes to plant structures, systems,
and components (SSCs), programmatic
controls, or the role of plant personnel
must include an assessment of the
changes in relation to the safety
requirements in subpart B of this part
and the analyses requirements of
§ 53.450 as applicable, an analysis of
whether the amendment involves no
significant hazards consideration using
the standards in § 53.1520, and a
consideration of environmental factors.
§ 53.1515 Public notices; State
consultation.

The Commission will use the
following procedures for an application
requesting an amendment to an OL or
COL issued under this part.
(a) Public notices. (1)(i) The
Commission may publish in the Federal
Register under § 2.105 of this chapter an
individual notice of proposed action for
an amendment for which it makes a
proposed determination that no
significant hazards consideration is
involved, or, at least once every 30 days,
publish a periodic Federal Register
notice of proposed actions, which
identifies each amendment issued and
each amendment proposed to be issued
since the last such periodic notice, or it
may publish both such notices.
(ii) For each amendment proposed to
be issued, the notice will
(A) Contain the staff’s proposed
determination under the standards in
§ 53.1520;
(B) Provide a brief description of the
amendment and of the facility involved;
(C) Solicit public comments on the
proposed determination; and
(D) Provide for a 30-day comment
period.
(iii) The comment period will begin
on the day after the date of the
publication of the first notice, and,
normally, the amendment will not be
granted until after this comment period
expires.

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(2) The Commission may inform the
public about the final disposition of an
amendment request for which it has
made a proposed determination of no
significant hazards consideration either
by issuing an individual notice of
issuance under § 2.106 of this chapter or
by publishing such a notice in its
periodic system of Federal Register
notices. In either event, it will not make
and will not publish a final
determination of no significant hazards
consideration unless it receives a
request for a hearing on that amendment
request.
(3) Where the Commission makes a
final determination that no significant
hazards consideration is involved and
that the amendment should be issued,
the amendment will be effective on
issuance, even if adverse public
comments have been received and even
if an interested person meeting the
provisions for intervention called for in
§ 2.309 of this chapter has filed a
request for a hearing. The Commission
need hold any required hearing only
after it issues an amendment, unless it
determines that a significant hazards
consideration is involved, in which case
the Commission will provide an
opportunity for a prior hearing.
(4) Where the Commission finds that
an emergency situation exists, in that
failure to act in a timely way would
result in derating or shutdown of a
commercial nuclear reactor, or in
prevention of either resumption of
operation or of increase in power output
up to the plant’s licensed power level,
it may issue a license amendment
involving no significant hazards
consideration without prior notice and
opportunity for a hearing or for public
comment. In such a situation, the
Commission will not publish a notice of
proposed determination on no
significant hazards consideration but
will publish a notice of issuance under
§ 2.106 of this chapter providing for
opportunity for a hearing and for public
comment after issuance. The
Commission expects its licensees to
apply for license amendments in a
timely fashion. It will decline to
dispense with notice and comment on
the determination of no significant
hazards consideration if it determines
that the licensee has abused the
emergency provision by failing to make
timely application for the amendment
and thus itself creating the emergency.
Whenever an emergency situation
exists, a licensee requesting an
amendment must explain why this
emergency situation occurred and why
it could not avoid this situation, and the
Commission will assess the licensee’s

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reasons for failing to file an application
sufficiently in advance of that event.
(5) Where the Commission finds that
exigent circumstances exist, in that a
licensee and the Commission must act
quickly and that time does not permit
the Commission to publish a Federal
Register notice allowing 30 days for
prior public comment, and it also
determines that the amendment
involves no significant hazards
considerations, it—
(i)(A) Will either issue a Federal
Register notice providing notice of an
opportunity for hearing and allowing at
least 2 weeks from the date of the notice
for prior public comment; or
(B) Will use local media to provide
reasonable notice to the public in the
area surrounding a licensee’s facility of
the licensee’s amendment and of its
proposed determination as described in
paragraph (a)(1) of this section,
consulting with the licensee on the
proposed media release and on the
geographical area of its coverage;
(ii) Will provide for a reasonable
opportunity for the public to comment,
using its best efforts to make available
to the public whatever means of
communication it can for the public to
respond quickly, and, in the case of
telephone comments, have these
comments recorded or transcribed, as
necessary and appropriate;
(iii) When it has issued a local media
release, may inform the licensee of the
public’s comments, as necessary and
appropriate;
(iv) Will publish a notice of issuance
under § 2.106 of this chapter;
(v) Will provide a hearing after
issuance, if one has been requested by
a person who satisfies the provisions for
intervention specified in § 2.309 of this
chapter; and
(vi) Will require the licensee to
explain the exigency and why the
licensee cannot avoid it and use its
normal public notice and comment
procedures in paragraph (a)(1) of this
section if it determines that the licensee
has failed to use its best efforts to make
a timely application for the amendment
in order to create the exigency and to
take advantage of this procedure.
(6) Where the Commission finds that
significant hazards considerations are
involved, it will issue a Federal Register
notice providing an opportunity for a
prior hearing even in an emergency
situation, unless it finds an imminent
danger to the health or safety of the
public, in which case it will issue an
appropriate order or rule under 10 CFR
part 2.
(b) State consultation. (1) At the time
a licensee requests an amendment, it
must notify the State in which its

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facility is located of its request by
providing that State with a copy of its
application and its reasoned analysis
about no significant hazards
considerations and indicate on the
application that it has done so.
(2) The Commission will advise the
State of its proposed determination
about no significant hazards
consideration normally by sending it a
copy of the Federal Register notice.
(3) The Commission will make the
names of the Project Manager or other
NRC personnel it designated to consult
with the State available to the State
official designated to consult about its
proposed determination. The
Commission will consider any
comments of that State official. If it does
not hear from the State in a timely
manner, it will consider that the State
has no interest in its determination;
nonetheless, to ensure that the State is
aware of the application, before it issues
the amendment, it will make a good
faith effort to communicate directly
with that official. (Inability to consult
with a responsible State official
following good faith attempts will not
prevent the Commission from making
effective a license amendment involving
no significant hazards consideration.)
(4) The Commission will make a good
faith attempt to consult with the State
before it issues a license amendment
involving no significant hazards
consideration. If, however, it does not
have time to use its normal consultation
procedures because of an emergency
situation, it will attempt to
communicate directly with the
appropriate State official. (Inability to
consult with a responsible State official
following good faith attempts will not
prevent the Commission from making
effective a license amendment involving
no significant hazards consideration, if
the Commission deems it necessary in
an emergency situation.)
(5) After the Commission issues the
requested amendment, it will send a
copy of its determination to the State.
(c) Caveats about State consultation.
(1) The State consultation procedures in
paragraph (b) of this section do not give
the State a right—
(i) To veto the Commission’s
proposed or final determination;
(ii) To a hearing on the determination
before the amendment becomes
effective; or
(iii) To insist upon a postponement of
the determination or upon issuance of
the amendment.
(2) These procedures do not alter
present provisions of law that reserve to
the Commission exclusive responsibility
for setting and enforcing radiological

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health and safety requirements for
commercial nuclear plants.
§ 53.1520

Issuance of amendment.

(a) In determining whether an
amendment to a license will be issued
to the applicant, the Commission will be
guided by the considerations which
govern the issuance of initial licenses to
the extent applicable and appropriate. If
the application is for amendment of an
OL or COL and involves the material
alteration of a commercial nuclear plant,
a construction permit (CP) will be
issued before the issuance of the
amendment to the license, provided
however, that if the application involves
a material alteration to a manufactured
reactor under this part before its
installation at a site, or a COL before the
date that the Commission makes the
finding under § 53.1452(g), no
application for or issuance of a CP is
required. If the amendment involves a
significant hazards consideration, the
Commission will give notice of its
proposed action—
(1) Under § 2.105 of this chapter
before acting thereon; and
(2) As soon as practicable after the
application has been docketed.
(b) The Commission will be
particularly sensitive to a license
amendment request that involves
irreversible consequences (such as one
that permits a significant increase in the
amount of effluents or radiation emitted
by a commercial nuclear plant).
(c) The Commission may make a final
determination, under the procedures in
§ 53.1515, that a proposed amendment
to an OL or a COL for a commercial
nuclear plant under this part involves
no significant hazards consideration, if
operation of the plant in accordance
with the proposed amendment would
not—
(1) Involve a significant increase in
the probability or consequences of an
accident previously evaluated; or
(2) Create the possibility of a new or
different kind of an accident from any
accident previously evaluated; or
(3) Involve a significant reduction in
a margin of safety.

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§ 53.1525 Revising certification
information within a design certification
rule.

(a) A holder of an OL or COL who
references a design certification rule
issued under this part must request an
exemption if proposing to change one or
more elements of the certification
information. The Commission may grant
such a request only if it determines that
the exemption will comply with the
requirements of § 53.080 and that the
special circumstances outweigh any

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decrease in safety that may result from
the reduction in standardization caused
by the departure.
(b) The request for an exemption must
be included with any associated license
amendment request, which must be
requested and processed in accordance
with §§ 53.1510, 53.1515, and 53.1520.
(c) Licensees must evaluate changes to
the design as described in the Final
Safety Analysis Report (FSAR) not
involving changes to the certification
information using the criteria in
§ 53.1550.
§ 53.1530 Revising design information
within a manufacturing license.

(a) The holder of an ML may not make
changes to the design of the
manufactured reactor authorized to be
manufactured without obtaining an
amendment pursuant to § 53.1510 and,
as applicable, § 53.1520.
(b) The holder of a COL under this
part who references or uses a
manufactured reactor under this part
must request approval for any proposed
departure from the design
characteristics, site parameters, terms
and conditions, or approved design of
the manufactured reactor. The
application for such departures must be
submitted and processed in accordance
with §§ 53.1510, 53.1515, and 53.1520.
In those cases where an ML references
a design certification rule, the
amendment application from the holder
of the COL must also request an
exemption from the design certification
rule under § 53.1525 if one or more
elements of the certification information
are adversely affected by the proposed
change. The holder of the COL must
evaluate changes to the commercial
nuclear plant as described in the FSAR
but outside of the scope of the
referenced ML using the criteria in
§ 53.1550.
§ 53.1535 Amendments during
construction.

(a) The holder of a CP or limited work
authorization (LWA) under this part
may request an amendment to the CP or
LWA in order to gain Commission
approval of the safety of selected design
features or specifications, including
proposed departures from a design
certification rule or ML. Amendments to
CPs or LWAs under this part must be
requested and processed under
§§ 53.1510 and 53.1520.
(b) The holder of a COL under this
part for which the NRC has not yet
made a finding in accordance with
§ 53.1452(g) must request amendments
required by § 53.1525 or § 53.1550 no
later than 45 days from the date the
licensee begins the construction of the

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SSCs to implement the change or
departure requiring NRC approval. The
licensee proceeds with such changes at
its own risk recognizing that there is a
possibility that the amendment will not
be granted.
§ 53.1540 Updating licensing-basis
information and determining the need for
NRC approval.

(a) Sections 53.1545 through 53.1565
provide the process for a holder of an
OL or COL to modify licensing-basis
information and to evaluate potential
changes to its facilities, procedures,
programs, and organizations to
determine if NRC approval is required.
(b) Definitions for the purposes of
§§ 53.1545 through 53.1565—
Change means a modification or
addition to, or removal from, the
commercial nuclear plant or procedures
that affects a design feature or related
functional design criteria, method of
performing or controlling the functions
of design features, or an evaluation that
demonstrates that intended functions
will be accomplished.
Departure from a method of
evaluation described in the Final Safety
Analysis Report (FSAR) (as updated)
used in establishing the functional
design criteria for safety-related
structures, systems, or components or in
the safety analyses means—
(1) Changing any of the elements of
the method described in the FSAR (as
updated) unless the results of the
analysis are conservative or essentially
the same; or
(2) Changing from a method described
in the FSAR to another method unless
that method has been approved by NRC
for the intended application.
Facility as described in the FSAR (as
updated) means—
(1) The SSCs that are described in the
FSAR (as updated),
(2) The design and performance
requirements for such SSCs described in
the FSAR (as updated), and
(3) The evaluations or methods of
evaluation included in the FSAR (as
updated) for such SSCs which
demonstrate that their intended
function(s) will be accomplished.
Final Safety Analysis Report (as
updated) means the FSAR submitted
under § 53.1369 or § 53.1416, as
amended and supplemented, and as
updated under § 53.1545, as applicable.
Procedures as described in the Final
Safety Analysis Report (as updated)
means those procedures that contain
information described in the FSAR (as
updated) such as how SSCs are operated
and controlled (including assumed
operator actions and response times).

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§ 53.1545
Reports.

Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules
Updating Final Safety Analysis

(a) Each holder of an OL or COL
under this part for which the
Commission has made the finding under
§ 53.1452(g) must update the FSAR
originally submitted as part of the
application for the license every 24
months or more frequently to assure that
the information included in the report
contains the latest information
developed. The submittal must include
the effects on the content of the FSAR
of—
(1) Changes made to the facility or
procedures as described in the FSAR;
(2) Safety analyses and evaluations
performed by the licensee either in
support of approved license
amendments or in support of
conclusions that changes did not require
a license amendment under § 53.1550;
(3) Updates to the probabilistic risk
assessments required under § 53.450;
(4) The cumulative effects of the
changes to the facility or procedures on
the margins to the safety criteria in
§§ 53.210, 53.220, 53.450(e), and 53.470
since the last FSAR update; and
(5) Analyses of new safety issues
performed by or on behalf of the
licensee at Commission request.
(b)(1) The licensee must submit
revisions containing updated
information to the Commission, under
§ 53.040, identifying the location of
revised or new information.
(2) The submittal must include—
(i) A certification by a duly authorized
officer of the licensee that either the
information accurately presents changes
made since the previous submittal,
necessary to reflect information and
analyses submitted to the Commission
or prepared pursuant to Commission
requirement, or that no such changes
were made; and
(ii) An identification of changes made
under the provisions of § 53.1550 but
not previously submitted to the
Commission.
(c) Each applicant for or holder of a
COL under this part for which the
Commission has not made the finding
under § 53.1452(g) must submit an
update to the FSAR annually by
providing the information required in
(a)(1) through (a)(5) of this section and
meeting the requirements of paragraph
(b) of this section. Combined license
applicants who have requested the NRC
to suspend its review of the COL
application and COL holders who have
informed the NRC that they do not plan
to pursue construction need not submit
an annual update of the FSAR. If a COL
applicant requests that the NRC resume
its review, or a COL holder notifies the
NRC that the COL holder plans to

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commence or resume construction, then
the COL applicant or holder must
submit to NRC an update to its FSAR
within 90 days of the request or
notification, as applicable, and annually
thereafter.
(d) The FSAR (as updated) must be
retained by the licensee until the
Commission terminates its license.
(e) Each holder of an ML under this
part must submit an update of the FSAR
reflecting any modification to the design
that is directed or approved by the
Commission under § 53.1288 or
§ 53.1530, and any new analyses of the
design requested by the Commission
under § 53.1580.
§ 53.1550 Evaluating changes to facility as
described in Final Safety Analysis Reports.

(a) The holder of an OL or COL may
make changes in the facility as
described in the FSAR (as updated) and
make changes in the procedures as
described in the FSAR (as updated)
without obtaining a license amendment
pursuant to § 53.1510 only if—
(1) A change to the technical
specifications incorporated in the
license is not required; and
(2) The change meets all of the
following criteria:
(i) Does not result in an increase to
the frequency or consequences of an
event sequence such that an event
sequence not previously identified as
risk significant becomes risk significant
by the analyses performed in
accordance with § 53.450(e).
(ii) Does not result in an increase to
the frequency or consequences of an
event sequence such that an event
sequence identified as risk significant in
accordance with § 53.450(e) exceeds the
licensing-basis event evaluation criteria
required to be established in accordance
with § 53.450(e).
(iii) Does not involve either of the
following: (A) a change to the NRCapproved comprehensive risk metric(s)
or associated risk performance objective
under § 53.220(b), or (B) an increase to
the frequency or consequences of one or
more event sequences such that there is
more than a minimal reduction in the
margin between the calculated
comprehensive risks posed by the
commercial nuclear plant and the safety
criteria of § 53.220.
(iv) Does not involve a departure from
a method of evaluation described in the
FSAR (as updated) used in assessing
licensing-basis events in accordance
with § 53.450 unless the results of the
analysis under § 53.450 are conservative
or essentially the same, the revised
method of evaluation has been
previously approved by the NRC for the
intended application, or the revised

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method of evaluation can be used under
an NRC-endorsed consensus code or
standard.
(v) Does not result in the escalation in
the safety classification of an SSC from
non-safety-related to non-safety-related
but safety-significant or from non-safetyrelated but safety-significant to safetyrelated.
(vi) Does not result in more than a
minimal decrease in defense in depth.
(vii) For commercial nuclear plants
licensed under this part for which
alternative evaluation criteria are
adopted in accordance with § 53.470,
does not result in a change to the
frequency or consequences of event
sequences such that the calculated
margins between the results for event
sequences evaluated in accordance with
§ 53.450(e) and the alternative
evaluation criteria decreases by 25
percent or more.
(viii) Does not result in the
identification of a new design-basis
accident in accordance with § 53.450(f).
(ix) Does not result in a decrease by
10 percent or more in the margin
between the consequence of any designbasis accident and the safety criteria in
§ 53.210.
(x) Does not prevent meeting the
design requirements in § 53.440(j) to
limit the release of radionuclides from
reactor systems, waste stores, or other
significant inventories of radioactive
materials assuming the impact of a
large, commercial aircraft.
(3) In implementing this paragraph,
the FSAR (as updated) is considered to
include FSAR changes since submittal
of the last update of the FSAR under
§ 53.1545.
(4) The provisions in this section do
not apply to changes to the facility or
procedures when the applicable
regulations establish more specific
criteria for accomplishing such changes.
(b)(1) A licensee who references a
design certification rule may make
departures from the standard design,
without prior Commission approval,
unless the proposed departure involves
a change to the design as described in
the rule certifying the design, in which
case the requirements of § 53.1525 are
applicable.
(2) The licensee must maintain
records of all departures from the
certified design of the facility and these
records must be maintained and
available for audit until the termination
of the license. The licensee must
identify the location and nature of
departures from licensing-basis
information within supporting
documents for a certified design within
the updates to the Safety Analysis
Report required by § 53.1545.

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(3) Licensees for which the NRC has
docketed the certifications required
under § 53.1070 need not retain records
of departures from the design of the
facility associated with SSCs that have
been permanently removed from service
using an NRC-approved change process.
(c)(1) The licensee must maintain
records of changes in the facility and
procedures made under paragraph (a) of
this section. These records must include
a written evaluation which provides the
bases for the determination that the
change does not require a license
amendment under paragraph (a)(2) of
this section.
(2) The licensee must submit, as
specified in § 53.040, a report
containing a brief description of any
departures and changes, including a
summary of the evaluation of each. A
report must be submitted at intervals
not to exceed 24 months. For COLs, the
report must be submitted at intervals
not to exceed 6 months during the
period from the date of application for
a COL to the date the Commission
makes its findings under § 53.1452(g).
(3) The records of changes in the
facility must be maintained until the
termination of an OL or COL issued
under this part, or the termination of a
renewed license issued under
§ 53.1595—whichever is later. Records
of changes in procedures must be
maintained for a period of 5 years.

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§ 53.1560 Updating program documents
included in licensing-basis information.

(a) Each holder under this part of an
OL or COL for which the Commission
has made the finding under § 53.1452(g)
must biennially or more frequently
update the program documents
submitted as part of an application to
obtain or maintain the license to assure
that the information included in the
documents contains the latest
information developed. The submittals
must include the effects on the content
of the program documents of—
(1) Changes made in the facility,
procedures, licensee’s organization, or
site environs;
(2) Safety analyses and evaluations
performed by the applicant or licensee
either in support of approved license
amendments or in support of
conclusions that changes did not require
a license amendment in accordance
with § 53.1550;
(3) Analyses of new safety issues
performed by or on behalf of the
licensee at Commission request; and
(4) Changes to the programs as a result
of operating experience, corrective
actions, or other reasons deemed
appropriate to ensure the programs
serve their underlying purpose to

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support the requirements in subpart B of
this part or other NRC regulations.
(b)(1) The licensee must submit
revisions containing updated
information to the Commission, as
specified in § 53.040, identifying the
location of revised or new information.
(2) The submittal must include—
(i) A certification by a duly authorized
officer of the licensee that either the
information accurately presents changes
made since the previous submittals,
necessary to reflect information and
analyses submitted to the Commission
or prepared pursuant to Commission
requirement, or that no such changes
were made; and
(ii) An identification of changes made
under the provisions of § 53.1550 but
not previously submitted to the
Commission.
(c) The updated program documents
must be retained by the licensee until
the Commission terminates their
license.
§ 53.1565 Evaluating changes to programs
included in licensing-basis information.

(a) A licensee may make changes to
the facility, procedures, or organizations
or address changes to site environs as
described in the program documents
included in licensing-basis information
without obtaining prior NRC approval
only if—
(1) A change to the technical
specifications incorporated in the
license is not required;
(2) An exemption from an NRC
regulation is not required; and
(3) The change conforms to programspecific requirements included in
regulations in this part, technical
specifications, or the NRC-approved
program document included and
reviewed as part of a license application
under subpart H or an amendment
under this subpart.
(b) In implementing this section, the
program documents (as updated)
include changes since submittal of the
last updates of the program documents
pursuant to § 53.1560.
(c) The provisions in this section do
not apply to changes to the program
documents when the applicable
regulations establish more specific
criteria for accomplishing such changes.
(d) To make changes to the facility,
procedures, or organizations or to
address changes to site environs as
described in the program documents
included in licensing-basis information
for individual programs, the following
requirements must be satisfied:
(1) Quality assurance program—
operation. (i) Each holder under this
part of an OL or COL, after the
Commission makes the finding under

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§ 53.1452(g), may make a change to a
previously accepted quality assurance
program (QAP) description included or
referenced in the Safety Analysis Report
without prior NRC approval, provided
the change does not reduce the
commitments in the program
description as accepted by the NRC.
Changes to the QAP description that do
not reduce the commitments must be
submitted to the NRC in accordance
with the requirements of § 53.1545. In
addition to QAP changes involving
administrative improvements and
clarifications, spelling corrections,
punctuation, or editorial items, the
following changes are not considered to
be reductions in commitment:
(A) The use of a quality assurance
(QA) standard approved by the NRC
which is more recent than the QA
standard in the licensee’s QAP at the
time of the change;
(B) The use of a QA alternative or
exception approved by an NRC safety
evaluation, provided that the bases of
the NRC approval are applicable to the
licensee’s facility;
(C) The use of generic organizational
position titles that clearly denote the
position function, supplemented as
necessary by descriptive text, rather
than specific titles;
(D) The use of generic organizational
charts to indicate functional
relationships, authorities, and
responsibilities, or, alternately, the use
of descriptive text;
(E) The elimination of QAP
information that duplicates language in
QA regulatory guides and QA standards
to which the licensee is committed; and
(F) Organizational revisions that
ensure that persons and organizations
performing QA functions continue to
have the requisite authority and
organizational freedom, including
sufficient independence from cost and
schedule when opposed to safety
considerations.
(ii) Changes to the QAP description
that do reduce the commitments must
be submitted to the NRC and receive
NRC approval prior to implementation,
as follows:
(A) Changes made to the QAP
description as presented in the Safety
Analysis Report or in a topical report
must be submitted as specified in
§ 53.040.
(B) The submittal of a change to the
Safety Analysis Report QAP description
must include all pages affected by that
change and must be accompanied by a
forwarding letter identifying the change,
the reason for the change, and the basis
for concluding that the revised program
incorporating the change continues to
satisfy the criteria of appendix B to part

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50 of this chapter and the Safety
Analysis Report QAP description
commitments previously accepted by
the NRC (the letter need not provide the
basis for changes that correct spelling,
punctuation, or editorial items).
(C) A copy of the forwarding letter
identifying the change must be
maintained as a facility record for 3
years.
(D) Changes to the QAP description
included or referenced in the Safety
Analysis Report shall be regarded as
accepted by the Commission upon
receipt of a letter to this effect from the
appropriate reviewing office of the
Commission or 60 days after submittal
to the Commission, whichever occurs
first.
(2) Quality assurance program—
siting, construction, and manufacturing.
Each holder of an LWA, early site
permit, CP, ML, or COL, before the
Commission makes the finding under
§ 53.1452(g) of this chapter, under this
part may make a change to a previously
accepted QAP description included or
referenced in the Safety Analysis Report
without prior NRC approval, provided
the change does not reduce the
commitments in the program
description previously accepted by the
NRC. Changes to the QAP description
that do not reduce the commitments
must be submitted to NRC within 90
days. Changes to the QAP description
that reduce the commitments must be
submitted to NRC and receive NRC
approval before implementation, as
follows:
(i) Changes to the Safety Analysis
Report must be submitted for review as
specified in § 53.040. Changes made to
NRC-accepted QA topical report
descriptions must be submitted as
specified in § 53.040.
(ii) The submittal of a change to the
Safety Analysis Report QAP description
must include all pages affected by that
change and must be accompanied by a
forwarding letter identifying the change,
the reason for the change, and the basis
for concluding that the revised program
incorporating the change continues to
satisfy the criteria of appendix B of part
50 of this chapter and the Safety
Analysis Report QAP description
commitments previously accepted by
the NRC (the letter need not provide the
basis for changes that correct spelling,
punctuation, or editorial items).
(iii) A copy of the forwarding letter
identifying the changes must be
maintained as a facility record for 3
years.
(iv) Changes to the QAP description
included or referenced in the Safety
Analysis Report shall be regarded as
accepted by the Commission upon

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receipt of a letter to this effect from the
appropriate reviewing office of the
Commission or 60 days after submittal
to the Commission, whichever occurs
first.
(3) Emergency preparedness program.
(i) Definitions for the purpose of
paragraph (d)(3) of this section:
(A) Change means an action that
results in modification or addition to, or
removal from, the licensee’s emergency
plan. All such changes are subject to the
provisions of this section except where
the applicable regulations establish
specific criteria for accomplishing a
particular change.
(B) Emergency plan means the
document(s), prepared and maintained
by the licensee, that identify and
describe the licensee’s methods for
maintaining emergency preparedness
and responding to emergencies. An
emergency plan includes the plan as
originally approved by the NRC and all
subsequent changes made by the
licensee with, and without, prior NRC
review and approval under paragraph
(d)(3) of this section.
(C) Emergency planning function
means a capability or resource necessary
to prepare for and respond to a
radiological emergency.
(D) Reduction in effectiveness means
a change in an emergency plan that
results in reducing the licensee’s
capability to perform an emergency
planning function in the event of a
radiological emergency.
(ii)(A) Except as provided in
paragraph (d)(3)(ii)(B) of this section, a
holder of an OL under this part, or a
COL under this part after the
Commission makes the finding under
§ 53.1452(g), must follow and maintain
the effectiveness of an emergency plan
that meets the requirements in appendix
E to part 50 of this chapter and the
planning standards of § 50.47(b).
(B) A holder of an OL under this part
for a commercial nuclear plant
consisting of small modular reactors
(SMRs) or non-light-water reactors, or a
holder of a COL under this part after the
Commission makes the finding under
§ 53.1452(g) for a commercial nuclear
plant consisting of either SMRs or nonlight-water reactors, must follow and
maintain the effectiveness of either an
emergency plan that meets the
requirements in § 50.160 or an
emergency plan that meets the
requirements in appendix E to part 50
of this chapter and the planning
standards of § 50.47(b).
(iii)(A) Except as provided in
paragraph (d)(3)(iii)(B) of this section,
the licensee may make changes to its
emergency plan without NRC approval
only if the licensee performs and retains

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an analysis demonstrating that the
changes do not reduce the effectiveness
of the plan and the plan, as changed,
continues to meet the requirements in
appendix E to part 50 of this chapter
and the planning standards of
§ 50.47(b).
(B) A license under this part for a
commercial nuclear plant consisting of
either SMRs or non-light-water reactors
may make changes to its emergency
plan without NRC approval only if the
licensee performs and retains an
analysis demonstrating that the changes
do not reduce the effectiveness of the
plan and the plan, as changed,
continues to meet either the
requirements in § 50.160 or the
requirements in appendix E to part 50
and the planning standards of
§ 50.47(b).
(iv) The changes to a licensee’s
emergency plan that reduce the
effectiveness of the plan as defined in
paragraph (d)(3)(i)(D) of this section
may not be implemented without prior
approval by the NRC. A licensee
desiring to make such a change must
submit an application for an
amendment to its license. In addition to
the filing requirements of §§ 53.1510
and 53.1515, the request must include
all emergency plan pages affected by
that change and must be accompanied
by a forwarding letter identifying the
change, the reason for the change, and
the basis for concluding that the
licensee’s emergency plan, as revised,
will continue to meet either the
requirements in § 50.160 to this chapter
or the requirements in appendix E to
part 50 of this chapter and the planning
standards of § 50.47(b) of this chapter.
(v) The licensee must retain a record
of each change to the emergency plan
made without prior NRC approval for a
period of three years from the date of
the change and shall submit, as
specified in § 53.040, a report of each
such change, including a summary of its
analysis, within 30 days after the change
is put in effect.
(vi) The licensee must retain the
emergency plan and each change for
which prior NRC approval was obtained
pursuant to paragraph (d)(3)(iv) of this
section as a record until the
Commission terminates the license for
the nuclear power reactor.
(vii)(A) The licensee must provide for
the development, revision,
implementation, and maintenance of its
emergency preparedness program. The
licensee must ensure that all program
elements are reviewed by persons who
have no direct responsibility for the
implementation of the emergency
preparedness program either—

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(1) At intervals not to exceed 12
months; or
(2) As necessary, based on an
assessment by the licensee against
performance indicators, and as soon as
reasonably practicable after a change
occurs in personnel, procedures,
equipment, or facilities that potentially
could adversely affect emergency
preparedness, but no longer than 12
months after the change. In any case, all
elements of the emergency preparedness
program must be reviewed at least once
every 24 months.
(B) The review must include an
evaluation for adequacy of interfaces
with State participating Tribal and local
governments and of licensee drills,
exercises, capabilities, and procedures.
The results of the review, along with
recommendations for improvements,
must be documented, reported to the
licensee’s corporate and plant
management, and retained for a period
of 5 years. The part of the review
involving the evaluation for adequacy of
interface with State, participating Tribal
and local governments must be available
to the appropriate State, participating
Tribal and local governments.
(4) Security programs. (i) The licensee
must prepare and maintain safeguards
contingency plan procedures in
accordance with appendix C of part 73
of this chapter for affecting the actions
and decisions contained in the
Responsibility Matrix of the safeguards
contingency plan. The licensee may not
make a change that would decrease the
safeguard effectiveness of a physical
security plan, or guard training and
qualification plan, or cybersecurity plan
submitted under subpart H or part 73 of
this chapter, or of the first four
categories of information (Background,
Generic Planning Base, Licensee
Planning Base, Responsibility Matrix)
contained in a licensee safeguards
contingency plan submitted under
subpart H or part 73 of this chapter, as
applicable, without prior approval of
the Commission. A licensee desiring to
make such a change must submit an
application for amendment to the
licensee’s license under §§ 53.1510,
53.1515, and 53.1520.
(ii) The licensee may make changes to
the plans referenced in paragraph (4)(i)
of this section without prior
Commission approval if the changes do
not decrease the safeguards
effectiveness of the plan. The licensee
must maintain records of changes to the
plans made without prior Commission
approval for a period of 3 years from the
date of the change, and must submit, as
specified in § 53.040, a report
containing a description of each change
within 2 months after the change is

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made. Prior to the safeguards
contingency plan being put into effect,
the licensee must have—
(A) All safeguards capabilities
specified in the safeguards contingency
plan available and functional;
(B) Detailed procedures developed
according to appendix C to part 73 of
this chapter available at the licensee’s
site; and
(C) All appropriate personnel trained
to respond to safeguards incidents as
outlined in the plan and specified in the
detailed procedures.
(iii) The licensee must provide for the
development, revision, implementation,
and maintenance of its safeguards
contingency plan. The licensee must
ensure that all program elements are
reviewed by individuals independent of
both security program management and
personnel who have direct
responsibility for implementation of the
security program either—
(A) At intervals not to exceed 12
months; or
(B) As necessary, based on an
assessment by the licensee against
performance indicators, and as soon as
reasonably practicable after a change
occurs in personnel, procedures,
equipment, or facilities that potentially
could adversely affect security, but no
longer than 12 months after the change.
In any case, all elements of the
safeguards contingency plan must be
reviewed at least once every 24 months.
(iv) The review must include a review
and audit of safeguards contingency
procedures and practices, an audit of
the security system testing and
maintenance program, and a test of the
safeguards systems along with
commitments established for response
by local law enforcement authorities.
The results of the review and audit,
along with recommendations for
improvements, must be documented,
reported to the licensee’s corporate and
plant management, and kept available at
the plant for inspection for a period of
3 years.
§ 53.1570

Transfer of licenses.

(a) No commercial nuclear plant
license issued under this part, or any
right thereunder, shall be transferred,
assigned, or in any manner disposed of,
either voluntarily or involuntarily,
directly or indirectly, through transfer of
control of the license to any person,
unless the Commission gives its consent
in writing.
(b)(1) An application for transfer of a
license must include—
(i) As much of the information
described in §§ 53.1109, 53.1306,
53.1366, and 53.1413 with respect to the
identity and technical and financial

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87109

qualifications of the proposed transferee
as would be required by those sections
if the application were for an initial
license. The Commission may require
additional information such as data
respecting proposed safeguards against
hazards from radioactive materials and
the applicant’s qualifications to protect
against such hazards.
(ii) A statement of the purposes for
which the transfer of the license is
requested, the nature of the transaction
necessitating or making desirable the
transfer of the license, and an agreement
to limit access to Restricted Data or
Classified National Security Information
pursuant to § 53.1115. The Commission
may require any person who submits an
application for license pursuant to the
provisions of this section to file a
written consent from the existing
licensee or a certified copy of an order
or judgment of a court of competent
jurisdiction attesting to the person’s
right (subject to the licensing
requirements of the Act and these
regulations) to possession of the facility
or site involved.
(2) [Reserved]
(c) After appropriate notice to
interested persons, including the
existing licensee, and observance of
such procedures as may be required by
the Act or regulations or orders of the
Commission, the Commission will
approve an application for the transfer
of a license, if the Commission
determines—
(1) That the proposed transferee is
qualified to be the holder of the license;
and
(2) That transfer of the license is
otherwise consistent with applicable
provisions of law, regulations, and
orders issued by the Commission
pursuant thereto.
§ 53.1575

Termination of licenses.

(a) When the holder of an OL or COL
under this part has determined to
permanently cease operations the
licensee must, within 30 days, submit a
written certification to the NRC,
consistent with the requirements of
§ 53.1070.
(b) Once fuel has been permanently
removed from the reactor system, the
licensee must submit a written
certification to the NRC that meets the
requirements of § 53.1070.
(c)(1) Upon docketing of the
certifications for permanent cessation of
operations and permanent removal of
fuel from the reactor system, or when a
final legally effective order to
permanently cease operations has come
into effect, the license no longer
authorizes operation of the reactor or

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emplacement or retention of fuel into
the reactor system.
(2) Activities associated with
decommissioning will be carried out in
accordance with the requirements and
procedures in subpart G of this part.
(3) The Commission shall terminate
the license if it determines that—
(i) The remaining dismantlement has
been performed in accordance with the
approved license termination plan
required in subpart G of this part; and
(ii) The final radiation survey and
associated documentation, including an
assessment of dose contributions
associated with parts released for use
before approval of the license
termination plan, demonstrate that the
facility and site have met the criteria for
decommissioning in subpart E of 10
CFR part 20.
(d) A holder of a CP or COL under this
part may request the termination of the
license as well as licenses issued by the
NRC under parts 30, 40, or 70 of this
chapter prior to plant operations. Such
requests may support an immediate
NRC approval of the site for unrestricted
use.
§ 53.1580

Information requests.

Each licensee under this part must at
any time before termination of the
license, upon request of the
Commission, submit, as specified in
§ 53.040 written statements, signed
under oath or affirmation, to enable the
Commission to determine whether or
not the license should be modified,
suspended, or revoked. Except for
information sought to verify licensee
compliance with the current licensing
basis for that facility, the NRC must
prepare the reason or reasons for each
information request prior to issuance to
ensure that the burden to be imposed on
respondents is justified in view of the
potential safety significance of the issue
to be addressed in the requested
information. Each such justification
provided for an evaluation performed by
the NRC staff must be approved by the
Executive Director for Operations or his
or her designee prior to issuance of the
request.

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§ 53.1585 Revocation, suspension,
modification of licenses and approvals for
cause.

A license or standard design approval
issued under this part may be revoked,
suspended, or modified, in whole or in
part, for any material false statement in
the application or in the supplemental
or other statement of fact required of the
applicant; or because of conditions
revealed by the application or statement
of fact of any report, record, inspection,
or other means which would warrant

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the Commission to refuse to grant a
license or approval on an original
application; or for failure to
manufacture a reactor, or construct or
operate a facility in accordance with the
terms of the license, provided, however,
that failure to make timely completion
of the proposed construction or
alteration of a facility under a CP under
this part shall be governed by the
provisions of § 53.1342(b); or for
violation of, or failure to observe, any of
the terms and provisions of the Act,
regulations, license, approval, or order
of the Commission.
§ 53.1590

Backfitting.

(a)(1) Backfitting means the
modification of or addition to systems,
structures, components, or design of a
facility; or the design approval or ML for
a facility; or the procedures or
organization required to design,
construct or operate a facility; any of
which may result from a new or
amended provision in the Commission’s
regulations or the imposition of a
regulatory staff position interpreting the
Commission’s regulations that is either
new or different from a previously
applicable staff position after the date of
the commercial nuclear plant license
issued under this part.
(2) Except as provided in paragraph
(a)(4) of this section, the Commission
shall require a systematic and
documented analysis pursuant to
paragraph (b) of this section for backfits
which it seeks to impose.
(3) Except as provided in paragraph
(a)(4) of this section, the Commission
shall require the backfitting of a facility
only when it determines, based on the
analysis described in paragraph (b) of
this section, that there is a substantial
increase in the overall protection of the
public health and safety or the common
defense and security to be derived from
the backfit and that the direct and
indirect costs of implementation for that
facility are justified in view of this
increased protection.
(4) The provisions of paragraphs (a)(2)
and (a)(3) of this section are
inapplicable and, therefore, backfit
analysis is not required and the
standards in paragraph (a)(3) of this
section do not apply where the
Commission or staff, as appropriate,
finds and declares, with appropriate
documented evaluation for its finding,
either—
(i) That a modification is necessary to
bring a facility into compliance with a
license or the rules or orders of the
Commission, or into conformance with
written commitments by the licensee; or
(ii) That regulatory action is necessary
to ensure that the facility provides

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adequate protection to the health and
safety of the public and is in accord
with the common defense and security;
or
(iii) That the regulatory action
involves defining or redefining what
level of protection to the public health
and safety or common defense and
security should be regarded as adequate.
(5) The Commission must always
require the backfitting of a facility if it
determines that such regulatory action
is necessary to ensure that the facility
provides adequate protection to the
health and safety of the public and is in
accord with the common defense and
security.
(6) The documented evaluation
required by paragraph (a)(4) of this
section must include a statement of the
objectives of and reasons for the
modification and the basis for invoking
the exception. If immediately effective
regulatory action is required, then the
documented evaluation may follow
rather than precede the regulatory
action.
(7) If there are two or more ways to
achieve compliance with a license or
the rules or orders of the Commission,
or with written licensee commitments,
or there are two or more ways to reach
a level of protection which is adequate,
then ordinarily the applicant or licensee
is free to choose the way which best
suits its purposes. However, should it be
necessary or appropriate for the
Commission to prescribe a specific way
to comply with its requirements or to
achieve adequate protection, then cost
may be a factor in selecting the way,
provided that the objective of
compliance or adequate protection is
met.
(b) In reaching the determination
required by paragraph (a)(3) of this
section, the Commission will consider
how the backfit should be scheduled in
light of other ongoing regulatory
activities at the facility and, in addition,
will consider information available
concerning any of the following factors
as may be appropriate and any other
information relevant and material to the
proposed backfit:
(1) The statement of the specific
objectives that the proposed backfit is
designed to achieve;
(2) The general description of the
activity that would be required by the
licensee or applicant in order to
complete the backfit;
(3) The potential change in the risk to
the public from the accidental off-site
release of radioactive material;
(4) The potential impact on
radiological exposure of facility
employees;

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(5) The installation and continuing
costs associated with the backfit,
including the cost of facility downtime
or the cost of construction delay;
(6) The potential safety impact of
changes in plant or operational
complexity, including the relationship
to proposed and existing regulatory
requirements;
(7) The estimated resource burden on
the NRC associated with the proposed
backfit and the availability of such
resources;
(8) The potential impact of differences
in facility type, design or age on the
relevancy and practicality of the
proposed backfit;
(9) Whether the proposed backfit is
interim or final and, if interim, the
justification for imposing the proposed
backfit on an interim basis.
(c) No licensing action will be
withheld during the pendency of backfit
analyses required by the Commission’s
rules.
(d) The Executive Director for
Operations shall be responsible for
implementation of this section, and all
analyses required by this section shall
be approved by the Executive Director
for Operations or his or her designee.
§ 53.1595

Renewal.

Licenses may be renewed by the
Commission upon expiration of the
period of the license.
Subpart J—Reporting and Other
Administrative Requirements
§ 53.1600

General information.

Each applicant and licensee under
this part must ensure that U.S. Nuclear
Regulatory Commission (NRC)
inspectors have unfettered access to
sites and facilities licensed or proposed
to be licensed in § 53.1610, must
maintain records and make reports to
the NRC in accordance with
requirements in §§ 53.1620 through
53.1650, must satisfy financial
qualification and reporting requirements
in §§ 53.1660 through 53.1700, and
must obtain and maintain required
financial protections in case of an
accident in §§ 53.1720 and 53.1730.

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§ 53.1610 Unfettered access for
inspections.

(a) Each applicant for or holder of a
manufacturing license (ML), operating
license (OL), combined license (COL),
construction permit (CP), or early site
permit must permit inspection, by duly
authorized representatives of the
Commission, of its records, premises,
activities, and of licensed materials in
possession or use, related to the license
or CP or early site permit as may be
necessary to effectuate the purposes of

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the Atomic Energy Act of 1956, as
amended, (the Act) and the Energy
Reorganization Act of 1974, as
amended.
(b)(1) Each holder of an ML, OL, COL,
or CP must, upon request by the
Director, Office of Nuclear Reactor
Regulation, provide rent-free office
space for the exclusive use of the
Commission inspection personnel. Heat,
air conditioning, light, electrical outlets,
and janitorial services must be
furnished by each licensee and each
holder of a CP. The office must be
convenient to and have full access to the
facility and must provide the inspectors
both visual and acoustic privacy.
(2) For a site or facility with an
assigned resident inspector, the space
provided must be adequate to
accommodate a full-time inspector, a
part-time secretary, and transient NRC
personnel and must be generally
commensurate with other office
facilities at the site. For sites or facilities
assigned multiple resident inspectors,
additional space may be requested. The
office space that is provided must be
subject to the approval of the Director,
Office of Nuclear Reactor Regulation.
All furniture, supplies, and
communication equipment will be
furnished by the Commission.
(3) For a site or facility without an
assigned resident inspector, temporary
space to accommodate periodic or
special inspections must be provided.
The office space must be generally
commensurate with other office
accommodations at the site.
(4) The licensee or permit holder must
afford any NRC resident inspector
assigned to that site, or other NRC
inspectors identified by the Regional
Administrator as likely to inspect the
facility, immediate unfettered access,
equivalent to access provided regular
plant employees, following proper
identification and compliance with
applicable access control measures for
security, radiological protection, and
personal safety.
(5) The licensee or permit holder must
ensure that the arrival and presence of
an NRC inspector, who has been
properly authorized facility access as
described in paragraph (b)(4) of this
section, is not announced or otherwise
communicated by its employees or
contractors to other persons at the
facility unless specifically requested by
the NRC inspector.
§ 53.1620 Maintenance of records, making
of reports.

(a) Each holder of an ML, OL, COL,
CP, or early site permit must maintain
all records and make all reports, in
connection with the activity, as may be

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required by the conditions of the license
or permit or by the regulations and
orders of the Commission in effectuating
the purposes of the Act and the Energy
Reorganization Act of 1974, as
amended. Reports must be submitted in
accordance with § 53.040.
(b) [Reserved]
(c) Records that are required by the
regulations in this part, by license
condition, or by technical specifications
must be retained for the period specified
by the appropriate regulation, license
condition, or technical specification. If
a retention period is not otherwise
specified, these records must be
retained until the Commission
terminates the facility license or, in the
case of an early site permit, until the
permit expires.
(d)(1) Records which must be retained
under this part may be the original or a
reproduced copy or a microform if the
reproduced copy or microform is duly
authenticated by authorized personnel
and the microform is capable of
producing a clear and legible copy after
storage for the period specified by
Commission regulations. The record
may also be stored in electronic media
with the capability of producing legible,
accurate, and complete records during
the required retention period. Records
such as letters, drawings, and
specifications, must include all
pertinent information such as stamps,
initials, and signatures. The licensee
must maintain adequate safeguards
against tampering with, and loss of
records.
(2) If there is a conflict between the
Commission’s regulations in this part,
license condition, or technical
specification, or other written
Commission approval or authorization
pertaining to the retention period for the
same type of record, the retention
period specified in the regulations in
this part for such records shall apply
unless the Commission, under § 53.080
of this part, has granted a specific
exemption from the record retention
requirements in the regulations in this
part.
(e) Each licensee must notify the
Commission as specified in § 53.040 of
this part, of successfully completing
power ascension testing or startup
testing as applicable within 30 calendar
days of completing the testing.
§ 53.1630 Immediate notification
requirements for operating commercial
nuclear plants.

(a) General requirements.1 (1) Each
holder of an OL under this part or a COL
under this part after the Commission
makes the finding under § 53.1452(g),
must notify the NRC Operations Center

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via the Emergency Notification System
(ENS) of—
(i) The declaration of any of the
Emergency Classes specified in the
licensee’s approved Emergency Plan; or
(ii) Those non-emergency events
specified in paragraph (b) of this section
that occurred within 3 years of the date
of discovery.
(2) If the ENS is inoperative, the
licensee must make the required
notifications via commercial telephone
service, other dedicated telephone
system, or any other method which will
ensure that a report is made as soon as
practical to the NRC Headquarters
Operations Center at the numbers
specified in appendix A to part 73 of
this chapter.
(3) The licensee must notify the NRC
immediately after notification of the
appropriate State or local agencies and
not later than 1 hour after the time the
licensee declares one of the Emergency
Classes.
(4) The licensee must activate the data
links with the NRC as specified in their
emergency plans after declaring an
Emergency Class for events of actual or
potential substantial degradation of
plant safety or security, probable risk to
site personnel life, or site equipment
damage caused by hostile action. The
data links may also be activated by the
licensee during emergency drills or
exercises if the licensee’s computer
system has the capability to transmit the
exercise data.
(5) When making a report under
paragraph (a)(1) of this section, the
licensee must identify—
(i) The Emergency Class declared; or
(ii) Paragraph (b)(1), ‘‘One-hour
reports,’’ paragraph (b)(2), ‘‘Four-hour
reports,’’ or paragraph (b)(3), ‘‘Eighthour reports,’’ as the paragraph of this
section requiring notification of the nonemergency event.
(b) Non-emergency events. (1) Onehour reports. If not reported as a
declaration of an Emergency Class
under paragraph (a) of this section, the
licensee must notify the NRC as soon as
practical and in all cases within one
hour of the occurrence of any deviation
from the plant’s Technical
Specifications authorized under
§ 53.740(h) of this part.
(2) Four-hour reports. If not reported
under paragraphs (a) or (b)(1) of this
section, the licensee must notify the
NRC as soon as practical, and in all
cases, within 4 hours of the occurrence
of any of the following:
(i) The initiation of any commercial
nuclear plant shutdown required by the
plant’s Technical Specifications.
(ii) Any event or condition that results
in actuation of the reactor protection

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system when the reactor is critical
except when the actuation results from
and is part of a pre-planned sequence
during testing or reactor operation.
(iii) Any event or condition that
results in an unplanned actuation of a
safety-related (SR) standby cooling
system or the unplanned sole reliance
on an SR standby cooling system for
those systems that are in constant
operation.
(iv) Any event or condition that
results in an unplanned movement of,
change of state in, or chemical
interaction involving a significant
amount of radioactive material within
the commercial nuclear plant.
(v) Any event or situation, related to
the health and safety of the public or
onsite personnel, or protection of the
environment, for which a news release
is planned or notification to other
government agencies has been or will be
made. Such an event may include an
onsite fatality or inadvertent release of
radioactively contaminated materials.
(3) Eight-hour reports. If not reported
under paragraphs (a), (b)(1), or (b)(2) of
this section, the licensee must notify the
NRC as soon as practical and in all cases
within 8 hours of the occurrence of any
of the following:
(i) Any event or condition that results
in—
(A) The condition of the commercial
nuclear plant, including its principal
safety barriers, being seriously
degraded; or
(B) The commercial nuclear plant
being in a condition not analyzed under
§ 53.450 that significantly degrades
plant safety.
(ii) Any event or condition that results
in valid actuation of an SR system,
except when the actuation results from
and is part of a pre-planned sequence
during testing or reactor operation.
(iii) Any event or condition that at the
time of discovery could have prevented
the fulfilment of the safety functions
identified under § 53.230. Events
covered may include one or more
procedural errors, equipment failures,
and/or discovery of design, analysis,
fabrication, construction, and/or
procedural inadequacies. However,
individual component failures need not
be reported pursuant to this paragraph
if other equipment was operable and
available to perform the required safety
function.
(iv) Any event requiring the transport
of a radioactively contaminated person
to an offsite medical facility for
treatment.
(v) Any event that results in a major
loss of emergency assessment capability,
offsite response capability, or offsite
communications capability (e.g.,

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significant portion of control room
indication, ENS, or offsite notification
system).
(c) Follow-up notification: With
respect to the notifications made under
paragraphs (a) and (b) of this section, in
addition to making the required initial
notification, each licensee, must during
the course of the event—
(1) Immediately Report:
(i) any further degradation in the level
of safety of the plant or other worsening
plant conditions, including those that
require the declaration of any of the
Emergency Classes, if such a declaration
has not been previously made, or
(ii) any change from one Emergency
Class to another, or
(iii) a termination of the Emergency
Class.
(2) Immediately Report:
(i) the results of ensuing evaluations
or assessments of plant conditions,
(ii) the effectiveness of response or
protective measures taken, and
(iii) important information related to
plant behavior that is not understood.
(3) Maintain an open, continuous
communication channel with the NRC
Operation Center upon request by the
NRC.
1 Other requirements for immediate
notification of the NRC by licensed operating
commercial nuclear plants are contained
elsewhere in this chapter, in particular
§§ 20.1906, 20.2202, 72.216, 73.77, and
73.1200 of this chapter.

§ 53.1640

Licensee event report system.

(a) Reportable events. (1) Each
commercial nuclear plant licensee
holding an OL under this part or a COL
under this part after the Commission
makes the finding under § 53.1452(g),
must submit a Licensee Event Report
(LER) for any event of the type
described in this paragraph within 60
days after discovery of the event. In the
case of an invalid actuation reported
under § 53.1640(a)(2), other than
automatic reactor shutdown when the
reactor is critical, the licensee may, at
its option, provide a telephone
notification to the NRC Operations
Center within 60 days after discovery of
the event instead of submitting a written
LER. Unless otherwise specified in this
section, the licensee must report an
event if it occurred within 3 years of the
date of discovery regardless of the plant
mode or power level, and regardless of
the significance of the structure, system,
or component that initiated the event.
(2) The licensee must report—
(i)(A) The completion of any
commercial nuclear plant shutdown
required by the plant’s Technical
Specifications.

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(B) Any operation or condition which
was prohibited by the plant’s Technical
Specifications except when—
(1) The Technical Specification is
administrative in nature;
(2) The event consisted solely of a
case of a late surveillance test where the
oversight was corrected, the test was
performed, and the equipment was
found to be capable of performing its
specified safety functions; or
(3) The Technical Specification was
revised prior to discovery of the event
such that the operation or condition was
no longer prohibited at the time of the
event.
(C) Any deviation from the plant’s
Technical Specifications authorized
under § 53.740(h).
(ii) Any event or condition that
resulted in—
(A) The condition of the commercial
nuclear plant, including its principal
safety barriers, being seriously
degraded; or
(B) The commercial nuclear plant
being in a condition not analyzed under
§ 53.450 that significantly degrades
plant safety.
(iii) Any natural phenomena or other
external condition that posed an actual
threat to the safety of the commercial
nuclear plant or significantly hampered
site personnel in the performance of
duties necessary for the safe operation
of the commercial nuclear plant.
(iv) Any event or condition that
resulted in inadvertent operation of any
structures, systems, and component
classified as SR for an identified safety
function under § 53.460 or the
unplanned sole reliance on an SR
system for those systems that are in
constant operation, except when—
(A) The actuation resulted from and
was part of a pre-planned sequence
during testing; or
(B) The actuation was invalid and—
(1) Occurred while the system was
properly removed from service; or
(2) Occurred after the safety function
had been already completed.
(v) Any event or condition that could
have prevented the fulfillment of the
safety functions identified under
§ 53.230.
(vi) Events covered in paragraph
(a)(2)(v) of this section may include one
or more procedural errors, equipment
failures, and/or discovery of design,
fabrication, construction, and/or
procedural inadequacies. However,
individual component failures need not
be reported pursuant to paragraph
(a)(2)(v) of this section if any other
equipment was operable and available
to perform the required safety function.
(vii)(A) Any event or condition that as
a result of a single cause could have

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prevented the fulfillment of any of the
safety functions identified under
§ 53.230.
(B) Events covered in paragraph
(a)(2)(vii)(A) of this section may include
cases of procedural error, equipment
failure, and/or discovery of a design,
analysis, fabrication, construction,
and/or procedural inadequacy.
However, licensees are not required to
report an event pursuant to paragraph
(a)(2)(vii)(A) of this section if the event
results from—
(1) A shared dependency among
trains or channels that is a natural or
expected consequence of the approved
plant design; or
(2) Normal and expected wear or
degradation.
(viii)(A) Any airborne radioactive
release that, when averaged over a time
period of 1-hour, resulted in airborne
radionuclide concentrations in an
unrestricted area that exceeds 20 times
the applicable concentration limits
specified in appendix B to 10 CFR part
20, table 2, column 1.
(B) Any liquid effluent release that,
when averaged over a time period of
1-hour, exceeds 20 times the applicable
concentrations specified in appendix B
to 10 CFR part 20, table 2, column 2, at
the point of entry into the receiving
waters (i.e., unrestricted area) for all
radionuclides except tritium and
dissolved noble gases.
(ix) Any event that posed an actual
threat to the safety of the commercial
nuclear plant or significantly hampered
site personnel in the performance of
duties necessary for the safe operation
of the plant, including fires, toxic gas
releases, or radioactive releases.
(b) Contents. The LER must contain—
(1) A brief abstract describing the
major occurrences during the event,
including all component or system
failures that contributed to the event
and significant corrective action taken
or planned to prevent recurrence.
(2)(i) A clear, specific narrative
description of what occurred so that
knowledgeable readers conversant with
the design of commercial nuclear plants,
but not familiar with the details of a
particular plant, can understand the
complete event.
(ii) The narrative description must
include the following specific
information as appropriate for the
particular event:
(A) Plant operating conditions before
the event.
(B) Status of systems, structures, or
components that were inoperable at the
start of the event and that contributed to
the event.
(C) Dates and approximate time of the
occurrences.

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(D) The cause of each component or
system failure or personnel error, if
known.
(E) The failure mode, mechanism, and
effect of each failed component, if
known.
(F) [Reserved]
(G) For failures of components with
multiple functions, include a list of
systems or secondary functions that
were also affected.
(H) For failure that rendered a
component or system classified as SR or
non-safety-related but safety-significant
inoperable, an estimate of the elapsed
time from the discovery of the failure
until the component or system was
returned to service.
(I) The method of discovery of each
component or system failure or
procedural error.
(J) For each human performance
related root cause, the licensee must
discuss the cause(s) and circumstances.
(K) Automatically and manually
initiated safety system responses.
(L) The manufacturer and model
number (or other identification) of each
component that failed during the event.
(3) An assessment of the safety
consequences and implications of the
event. This assessment must include—
(i) The availability of systems or
components that could have performed
the same function as the components
and systems that failed during the event,
and
(ii) For events that occurred when the
reactor was shut down, the availability
of systems or components that are
needed to shut down the reactor and
maintain safe shutdown conditions,
remove residual heat, control the release
of radioactive material, or mitigate the
consequences of an accident.
(4) A description of any corrective
actions planned as a result of the event,
including those to reduce the
probability of similar events occurring
in the future.
(5) Reference to any previous similar
events at the same plant that are known
to the licensee.
(6) The name and contact information
of a person within the licensee’s
organization who is knowledgeable
about the event and can provide
additional information concerning the
event and the plant’s characteristics.
(c) Supplemental Information. The
Commission may require the licensee to
submit specific additional information
beyond that required by paragraph (b) of
this section if the Commission finds that
supplemental material is necessary for
complete understanding of an unusually
complex or significant event. These
requests for supplemental information
will be made in writing and the licensee

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must submit, as specified in § 53.040,
the requested information as a
supplement to the initial LER.
(d) Submission of Reports. Licensee
Event Reports must be prepared on
Form NRC 366 and submitted to the
NRC, as specified in § 53.040.
(e) Report Legibility. The reports and
copies that licensees are required to
submit to the Commission under the
provisions of this section must be of
sufficient quality to permit legible
reproduction and micrographic
processing.

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53.1645 Reports of radiation exposure to
members of the public.

(a) Each holder of an OL, and each
holder of a COL after the Commission
has made the finding under
§ 53.1452(g), must submit radiological
reports as required by 10 CFR part 20,
as well as an Annual Radioactive
Effluent Release Report and an Annual
Radiological Environmental Operating
Report. The Annual Radioactive
Effluent Release Report must specify the
quantity of each of the principal
radionuclides released to unrestricted
areas in liquid and in gaseous effluents
and an estimate of the dose received by
the maximally exposed member of the
public in an unrestricted area from
effluents and direct radiation from
contained sources during the previous
calendar year. The Annual Radiological
Environmental Operating Report must
provide data on measurable levels of
radiation and radioactive materials in
the environment, must include an
evaluation of the relationship between
quantities of radioactive material
released in effluents and resultant
radiation doses to individuals from
principal pathways of exposure, and
must include the results of
environmental monitoring during the
previous calendar year. These reports
must also include any other information
as may be required by the Commission
to estimate maximum potential annual
radiation doses to the public. The
reports must be submitted as specified
in § 53.040 by May 15 of each
successive year. If the total effective
dose equivalent to members of the
public in unrestricted areas during the
reporting period is greater than the as
low as is reasonably achievable
(ALARA) design objectives established
under § 53.425, the report must specify
the causes for exceeding the ALARA
design objective and describe any
corrective actions. On the basis of these
reports and any additional information
the Commission may obtain from the
licensee or others, the Commission may
require the licensee to take action as the
Commission deems appropriate.

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(b) If during any calendar quarter the
radiation exposure to a member of the
public in the unrestricted areas,
calculated on the same basis as the
respective ALARA design objective
exposure, exceeds one-half of the
annual ALARA design objective
exposure, the licensee must submit a
report as specified in § 53.040. The
report shall specify the causes for
exceeding one-half the annual ALARA
design objective exposure in a quarter
and describe corrective actions that the
licensee will take to maintain radiation
exposure to levels within the ALARA
design objectives for the remainder of
the year. The report shall be submitted
within 30 days from the end of the
quarter when one-half of the annual
ALARA design objective exposure was
exceeded.
§ 53.1650 Facility information and
verification.

(a) In response to a written request by
the Commission, each applicant for a CP
or license and each recipient of a CP or
a license must submit facility
information, as described in § 75.10 of
this chapter, on International Atomic
Energy Agency (IAEA) Design
Information Questionnaire forms and
site information on DOC/NRC Form AP–
A and associated forms;
(b) As required by the Additional
Protocol, must submit location
information described in § 75.11 of this
chapter on DOC/NRC Form AP–1 and
associated forms; and
(c) Must permit verification thereof by
the IAEA and take other action as
necessary to implement the US/IAEA
Safeguards Agreement, as described in
part 75 of this chapter.
§ 53.1660

Financial requirements.

Sections 53.1670 through 53.1700 set
out the requirements and procedures
related to financial qualifications and
related reporting requirements.
§ 53.1670

Financial qualifications.

Except for an electric utility applicant
for a license to operate a commercial
nuclear plant, an applicant for a CP, OL,
or COL under this part must possess or
have reasonable assurance of obtaining
the funds necessary for the activities for
which the permit or license is sought.
§ 53.1680

Annual financial reports.

With respect to any commercial
nuclear plant of a type described in
§ 53.020, each licensee and each holder
of a CP must submit its annual financial
report, including the certified financial
statements, to the Commission, as
specified in § 53.040, upon issuance of
the report. However, licensees and
holders of a CP who submit a Form 10–

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Q with the Securities and Exchange
Commission or a Form 1 with the
Federal Energy Regulatory Commission
need not submit the annual financial
report or the certified financial
statement under this section.
§ 53.1690 Licensee’s change of status;
financial qualifications.

(a) An electric utility licensee holding
an OL or COL (including a renewed
license) for a commercial nuclear plant,
no later than seventy-five (75) days prior
to ceasing to be an electric utility in any
manner not involving a license transfer
under § 53.1399 or § 53.1456 must
provide the NRC with the financial
qualifications information that would be
required for obtaining an initial OL or
COL under this part. The financial
qualifications information must address
the first full 5 years of operation after
the date the licensee ceases to be an
electric utility.
(b)(1) Any holder of a license issued
under this part must notify the
appropriate NRC Regional
Administrator, in writing, immediately
following the filing of a voluntary or
involuntary petition for bankruptcy
under any chapter of title 11
(Bankruptcy) of the United States Code
by or against—
(i) The licensee;
(ii) An entity (as 11 U.S.C. 101(14)
defines that term) controlling the
licensee or listing the license or licensee
as property of the estate; or
(iii) An affiliate (as 11 U.S.C. 101(2)
defines that term) of the licensee.
(2) This notification must indicate—
(i) The bankruptcy court in which the
petition for bankruptcy was filed; and
(ii) The date of the filing of the
petition.
§ 53.1700

Creditor regulations.

(a) Pursuant to section 184 of the Act,
the Commission consents, without
individual application, to the creation of
any mortgage, pledge, or other lien upon
any facility not owned by the United
States which is the subject of a license
or upon any leasehold or other interest
in such facility; provided—
(1) That the rights of any creditor so
secured may be exercised only in
compliance with and subject to the
same requirements and restrictions as
would apply to the licensee pursuant to
the provisions of the license, the Act,
and regulations issued by the
Commission under the Act; and
(2) That no creditor so secured may
take possession of the facility pursuant
to the provisions of this section prior to
either the issuance of a license from the
Commission authorizing such
possession or the transfer of the license.

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Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules
(b) Any creditor so secured may apply
for transfer of the license covering such
facility by filing an application for
transfer of the license under § 53.1570.
The Commission will act upon such
application under subpart I of this part.
(c) Nothing contained in this
regulation shall be deemed to affect the
means of acquiring, or the priority of,
any tax lien or other lien provided by
law.
(d) As used in this section—
License includes any license under
this part, which may be issued by the
Commission with regard to a facility.
Creditor includes, without implied
limitation, the trustee under any
mortgage, pledge or lien on a facility
made to secure any creditor, any trustee
or receiver of the facility appointed by
a court of competent jurisdiction in any
action brought for the benefit of any
creditor secured by such mortgage,
pledge or lien, any purchaser of such
facility at the sale thereof upon
foreclosure of such mortgage, pledge, or
lien or upon exercise of any power of
sale contained therein, or any assignee
of any such purchaser.
Facility includes, but is not limited to,
a site which is the subject of an early
site permit under this part, and a reactor
manufactured under an ML under this
part.
§ 53.1710

Financial protection.

Sections 53.1720 and 53.1730 set out
the requirements and procedures related
to licensees obtaining and maintaining
insurance to cover stabilization and
decontamination activities in the event
of an accident and financial protection
in accordance with part 140, ‘‘Financial
Protection Requirements and Indemnity
Agreements,’’ of this chapter.

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§ 53.1720 Insurance required to stabilize
and decontaminate plant following an
accident.

Each commercial nuclear plant
licensee under this part must take
reasonable steps to obtain insurance
available at reasonable costs and on
reasonable terms from private sources or
to demonstrate that it possesses an
equivalent amount of protection
covering the licensee’s obligation, in the
event of an accident at the licensee’s
commercial nuclear reactor, to stabilize
and decontaminate the plant and the
plant site at which such an accident
may occur, provided that—
(a) The insurance required by this
section must have a minimum coverage
limit for each commercial nuclear plant
site of $1.06 billion, an amount based on
plant-specific estimates of costs to
stabilize and decontaminate a plant, or
whatever amount of insurance is

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generally available from private sources,
whichever is less. The required
insurance must clearly state that, as and
to the extent provided in paragraph
(d)(1) of this section, any proceeds must
be payable first for stabilization of the
plant and next for decontamination of
the plant and the plant site. If a
licensee’s coverage falls below the
required minimum, the licensee must
within 60 days take all reasonable steps
to restore its coverage to the required
minimum. The required insurance may,
at the option of the licensee, be
included within policies that also
provide coverage for other risks,
including, but not limited to, the risk of
direct physical damage.
(b)(1) With respect to policies issued
or annually renewed, the proceeds of
such required insurance must be
dedicated, as and to the extent provided
in this paragraph, to reimbursement or
payment on behalf of the insured of
reasonable expenses incurred or
estimated to be incurred by the licensee
in taking action to fulfill the licensee’s
obligation, in the event of an accident at
the licensee’s plant, to ensure that the
plant is in, or is returned to, and
maintained in, a safe and stable
condition and that radioactive
contamination is removed or controlled
such that personnel exposures are
consistent with the occupational
exposure limits in 10 CFR part 20.
These actions must be consistent with
any other obligation the licensee may
have under this chapter and must be
subject to paragraph (d) of this section.
As used in this section, an ‘‘accident’’
means an event that involves the release
of radioactive material from its intended
place of confinement within the
commercial nuclear plant such that
there is a present danger of release off
site in amounts that would pose a threat
to the public health and safety.
(2) The stabilization and
decontamination requirements set forth
in paragraph (d) of this section must
apply uniformly to all insurance
policies required under this section.
(c) The licensee shall report to the
NRC on April 1 of each year the current
levels of this insurance or financial
security it maintains and the sources of
this insurance or financial security.
(d)(1) In the event of an accident at
the licensee’s plant, whenever the
estimated costs of stabilizing the
licensed plant and of decontaminating
the plant and the plant site exceed one
tenth of the minimum insurance under
paragraph (a) of this section, the
proceeds of the insurance required by
this section must be dedicated to and
used, first, to ensure that the licensed
plant is in, or is returned to, and can be

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maintained in, a safe and stable
condition so as to prevent any
significant risk to the public health and
safety and, second, to decontaminate the
plant and the plant site in accordance
with the licensee’s cleanup plan as
approved by order of the Director, Office
of Nuclear Reactor Regulation. This
priority on insurance proceeds must
remain in effect for 60 days or, upon
order of the Director, for such longer
periods, in increments not to exceed 60
days except as provided for activities
under the cleanup plan required in
paragraphs (d)(3) and (d)(4) of this
section, as the Director may find
necessary to protect the public health
and safety. Actions needed to bring the
plant to and maintain the plant in a safe
and stable condition may include one or
more of the following, as appropriate:
(i) Shutdown of the reactor(s) and
other processes at the plant;
(ii) Establishment and maintenance of
long-term cooling with stable decay heat
removal;
(iii) Maintenance of sub-criticality;
(iv) Control of radioactive releases;
and
(v) Securing of structures, systems, or
components to minimize radiation
exposure to onsite personnel or to the
offsite public or to facilitate later
decontamination or both.
(2) The licensee must inform the
Director, Office of Nuclear Reactor
Regulation in writing when the plant is
and can be maintained in a safe and
stable condition so as to prevent any
significant risk to the public health and
safety. Within 30 days after the licensee
informs the Director that the plant is in
this condition, or at such earlier time as
the licensee may elect or the Director
may for good cause direct, the licensee
must prepare and submit a cleanup plan
for the Director’s approval. The cleanup
plan must identify and contain an
estimate of the cost of each cleanup
operation that will be required to
decontaminate the reactor sufficiently to
permit the licensee either to resume
operation of the reactor or to apply to
the Commission under subpart G of this
part for authority to decommission the
reactor and to surrender the license
voluntarily. Cleanup operations may
include one or more of the following, as
appropriate:
(i) Processing any contaminated
materials generated by the accident and
by decontamination operations to
remove radioactive materials;
(ii) Decontamination of surfaces
inside the plant buildings to levels
consistent with the Commission’s
occupational exposure limits in 10 CFR
part 20, and decontamination or
disposal of equipment;

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Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules

(iii) Decontamination or removal and
disposal of internal parts, damaged fuel
from the reactor coolant or fuel systems,
or related process or waste systems; and
(iv) Cleanup of the reactor coolant or
fuel systems or related process or waste
systems.
(3) Following review of the licensee’s
cleanup plan, the Director will order the
licensee to complete all operations that
the Director finds are necessary to
decontaminate the reactor sufficiently to
permit the licensee either to resume
operation of the reactor or to apply to
the Commission under subpart G of this
part for authority to decommission the
reactor and to surrender the license
voluntarily. The Director must approve
or disapprove, in whole or in part for
stated reasons, the licensee’s estimate of
cleanup costs for such operations. Such
order may not be effective for more than
one year, at which time it may be
renewed. Each subsequent renewal
order, if imposed, may be effective for
not more than 6 months.
(4) Of the balance of the proceeds of
the required insurance not already
expended to place the plant in a safe
and stable condition under paragraph
(b)(1) of this section, an amount
sufficient to cover the expenses of
completion of those decontamination
operations that are the subject of the
Director’s order must be dedicated to
such use, provided that, upon
certification to the Director of the
amounts expended previously and from
time to time for stabilization and
decontamination and upon further
certification to the Director as to the
sufficiency of the dedicated amount
remaining, policies of insurance may
provide for payment to the licensee or
other loss payees of amounts not so
dedicated, and the licensee may proceed
to use in parallel (and not in preference
thereto) any insurance proceeds not so
dedicated for other purposes.
§ 53.1730 Financial protection
requirements.

Commercial nuclear plant licensees
must satisfy the applicable provisions of
part 140, ‘‘Financial Protection
Requirements and Indemnity
Agreements,’’ of this chapter.

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Subparts K and L [Reserved]
Subpart M—Enforcement
§ 53.9000

Violations.

(a) The Commission may obtain an
injunction or other court order to
prevent a violation of the provisions
of—
(1) The Atomic Energy Act of 1954, as
amended (the Act);

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(2) Title II of the Energy
Reorganization Act of 1974, as
amended; or
(3) A regulation or order issued under
those Acts.
(b) The Commission may obtain a
court order for the payment of a civil
penalty imposed under Section 234 of
the Act:
(1) For violations of—
(i) Sections 53, 57, 62, 63, 81, 82, 101,
103, 104, 107, or 109 of the Act;
(ii) Section 206 of the Energy
Reorganization Act of 1974, as
amended;
(iii) Any rule, regulation, or order
issued under the sections specified in
paragraph (b)(1)(i) of this section;
(iv) Any term, condition, or limitation
of any license issued under the sections
specified in paragraph (b)(1)(i) of this
section.
(2) For any violation for which a
license may be revoked under section
186 of the Act.
§ 53.9010

Criminal penalties.

(a) Section 223 of the Act provides for
criminal sanctions for willful violation
of, attempted violation of, or conspiracy
to violate, any regulation issued under
sections 161b, 161i, or 161o of the Act.
For purposes of section 223, all the
regulations in part 53 are issued under
one or more of sections 161b, 161i, or
161o, except for the sections listed in
paragraph (b) of this section.
(b) The regulations in 10 CFR part 53
that are not issued under sections 161b,
161i, or 161o for the purposes of section
223 are as follows: §§ 53.000, 53.015,
53.020, 53.040, 53.080, 53.090, 53.100,
53.110, 53.120, 53.600, 53.725, 53.726,
53.735, 53.760, 53.775, 53.790, 53.795,
53.820, 53.910, 53.1000, 53.1050,
53.1100, 53.1103, 53.1106, 53.1109,
53.1112, 53.1115, 53.1118, 53.1120,
53.1121, 53.1124, 53.1140, 53.1143,
53.1144, 53.1146, 53.1149, 53.1155,
53.1158, 53.1164, 53.1170, 53.1173,
53.1176, 53.1179, 53.1188, 53.1200,
53.1203, 53.1206, 53.1209, 53.1210,
53.1212, 53.1215, 53.1218, 53.1221,
53.1230, 53.1236, 53.1239, 53.1241,
53.1242, 53.1245, 53.1248, 53.1251,
53.1254, 52.1257, 52.1260, 53.1263,
53.1270, 53.1273, 53.1276, 53.1279,
53.1282, 53.1285, 53.1286, 53.1287,
53.1288, 53.1291, 53.1293, 53.125,
53.1300, 53.1306, 53.1309, 53.1312,
53.1315, 53.1318, 53.1324, 53.1330,
53.1333, 53.1336, 53.1348, 53.1360,
53.1366, 53.1369, 53.1372, 53.1375,
53.1381, 53.1384, 53.1387, 53.1390,
53.1396, 53.1401, 53.1405, 53.1410,
53.1416, 53.1419, 53.1422, 53.1425,
53.1431, 53.1437, 53.1440, 53.1443,
53.1452, 53.1455, 53.1456, 53.1458,
53.1461, 53.1470, 53.1500, 53.1510,

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53.1515, 53.1520, 53.1525, 53.1530,
53.1535, 53.1540, 53.1560, 53.1585,
53.1590, 53.1595, 53.1600, 53.1660,
53.1670, 53.1700, 53.1710, 53.1730,
53.9000, 53.9010.
PART 70—DOMESTIC LICENSING OF
SPECIAL NUCLEAR MATERIAL
129. The authority citation for 10 CFR
part 70 continues to read as follows:

■

Authority: Atomic Energy Act of 1954,
secs. 51, 53, 57(d), 108, 122, 161, 182, 183,
184, 186, 187, 193, 223, 234, 274, 1701 (42
U.S.C. 2071, 2073, 2077(d), 2138, 2152, 2201,
2232, 2233, 2234, 2236, 2237, 2243, 2273,
2282, 2021, 2297f); Energy Reorganization
Act of 1974, secs. 201, 202, 206, 211 (42
U.S.C. 5841, 5842, 5846, 5851); Nuclear
Waste Policy Act of 1982, secs. 135, 141 (42
U.S.C. 10155, 10161); 44 U.S.C. 3504 note.
§ 70.20a

[Amended]

130. In § 70.20a, in paragraph (b)
remove the phrase ‘‘parts 30 through 36,
39, 40, 50, 72, 110,’’ and add in its place
the phrase ‘‘parts 30 through 36, 39, 40,
50, 53, 72, 110’’.

■

§ 70.22

[Amended]

131. In § 70.22, wherever it appears,
remove the phrase ‘‘part 50’’ and add in
its place the phrase ‘‘parst 50 or 53’’.
■ 132. In § 70.24, revise paragraph (d) to
read as follows:
■

§ 70.24

Criticality accident requirements.

*

*
*
*
*
(d)(1) The requirements in paragraphs
(a) through (c) of this section do not
apply to a holder of a construction
permit or operating license for a nuclear
power reactor issued under part 50 or
part 53 of this chapter or a combined
license issued under part 52 or part 53
of this chapter, if the holder complies
with the requirements of paragraph (b)
of 10 CFR 50.68 or paragraph (m)(2) of
10 CFR 53.440,as applicable.
(2) An exemption from § 70.24 held
by a licensee who thereafter elects to
comply with requirements of paragraph
(b) of 10 CFR 50.68 or paragraph (m)(2)
of 10 CFR 53.440 does not exempt that
licensee from complying with any of the
requirements in § 50.68 or § 53.440(m)
of this chapter but shall be ineffective so
long as the licensee elects to comply
with § 50.68(b) or § 53.440(m)(2) of this
chapter, as applicable.
§ 70.32

[Amended]

133. In § 70.32, in paragraph (c)(1)
introductory text, remove the phrase
‘‘part 50 of this chapter’’ and add in its
place the phrase ‘‘parts 50 or 53 of this
chapter’’; and in paragraph (d) remove
the phrase ‘‘or § 70.34 of this chapter, as
appropriate.’’ and add in its place the
phrase ‘‘, §§ 74.34 or 53.1510 of this
chapter, as appropriate.’’.

■

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Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules
134. In § 70.50, revise paragraph (d) to
read as follows:

■

§ 70.50

Reporting requirements.

*

*
*
*
*
(d) The provisions of § 70.50 do not
apply to licensees subject to §§ 50.72 or
53.1630 of this chapter. They do apply
to those 10 CFR parts 50 or 53 licensees
possessing material licensed under 10
CFR part 70 that are not subject to the
notification requirements in §§ 50.72 or
53.1630 of this chapter.
PART 72—LICENSING
REQUIREMENTS FOR THE
INDEPENDENT STORAGE OF SPENT
NUCLEAR FUEL AND HIGH-LEVEL
RADIOACTIVE WASTE, AND
REACTOR-RELATED GREATER THAN
CLASS C WASTE
135. The authority citation for 10 CFR
part 72 continues to read as follows:

■

Authority: Atomic Energy Act of 1954,
secs. 51, 53, 57, 62, 63, 65, 69, 81, 161, 182,
183, 184, 186, 187, 189, 223, 234, 274 (42
U.S.C. 2071, 2073, 2077, 2092, 2093, 2095,
2099, 2111, 2201, 2210e, 2232, 2233, 2234,
2236, 2237, 2238, 2273, 2282, 2021); Energy
Reorganization Act of 1974, secs. 201, 202,
206, 211 (42 U.S.C. 5841, 5842, 5846, 5851);
National Environmental Policy Act of 1969
(42 U.S.C. 4332); Nuclear Waste Policy Act
of 1982, secs. 117(a), 132, 133, 134, 135, 137,
141, 145(g), 148, 218(a) (42 U.S.C. 10137(a),
10152, 10153, 10154, 10155, 10157, 10161,
10165(g), 10168, 10198(a)); 44 U.S.C. 3504
note.

136. In § 72.3, revise the definition for
‘‘Independent spent fuel storage
installation or ISFSI’’ to read as follows:

■

§ 72.3

Definitions.

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*

*
*
*
*
Independent spent fuel storage
installation or ISFSI means a complex
designed and constructed for the
interim storage of spent nuclear fuel,
solid reactor-related GTCC waste, and
other radioactive materials associated
with spent fuel and reactor-related
GTCC waste storage. An ISFSI which is
located on the site of another facility
licensed under this part or a facility
licensed under part 50 or part 53 of this
chapter and which shares common
utilities and services with that facility or
is physically connected with that other
facility may still be considered
independent.
*
*
*
*
*
■ 137. In § 72.30, revise paragraph (e)(5)
to read as follows:
§ 72.30 Financial assurance and
recordkeeping for decommissioning.

*

*
*
*
*
(e) * * *
(5) In the case of licensees who are
issued a power reactor license under

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parts 50 or 53 of this chapter or ISFSI
licensees who are an electric utility, as
defined in parts 50 or 53 of this chapter,
with a specific license issued under this
part, the methods of §§ 50.75(b), (e), and
(h) or 53.1010, 53.1040, 53.1045(b), and
53.1060 of this chapter, as applicable. In
the event that funds remaining to be
placed into the licensee’s ISFSI
decommissioning external sinking fund
are no longer approved for recovery in
rates by a competent rate making
authority, the licensee must make
changes to provide financial assurance
using one or more of the methods stated
in paragraphs (a)(1) through (4) of this
section.
*
*
*
*
*
■ 138. In § 72.32, revise paragraph (c)(2)
to read as follows:
§ 72.32

Emergency plan.

*

*
*
*
*
(c) * * *
(2)(i) Located within the exclusion
area as defined in 10 CFR part 100, of
a nuclear power reactor licensed for
operation by the Commission, the
emergency plan that meets either the
requirements in § 50.160 of this chapter
or the requirements in appendix E to
part 50 of this chapter and § 50.47(b) of
this chapter shall be deemed to satisfy
the requirements of this section.
(ii) Located within the exclusion area,
as defined in 10 CFR part 53, of a
commercial nuclear plant licensed for
operation by the Commission, the
emergency plan that meets either the
requirements in § 50.160 of this chapter
or the requirements in appendix E to
part 50 of this chapter and § 50.47(b) of
this chapter shall be deemed to satisfy
the requirements of this section.
*
*
*
*
*
§ 72.40

[Amended]

139. In § 72.40, in paragraph (c)
remove the phrase ‘‘under part 50 of this
chapter,’’ and add in its place the phrase
‘‘under parts 50 or 53 of this chapter,’’.
■ 140. In § 72.75, revise paragraph
(i)(1)(ii) to read as follows:
■

§ 72.75 Reporting requirements for
specific events and conditions.

*

*
*
*
*
(i) * * *
(1) * * *
(ii) Licensees issued a general license
under § 72.210, after the licensee has
placed spent fuel on the ISFSI storage
pad (if the ISFSI is located inside the
collocated protected area, for a reactor
licensed under parts 50 or 53 of this
chapter) or after the licensee has
transferred spent fuel waste outside the
reactor licensee’s protected area to the
ISFSI storage pad (if the ISFSI is located

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87117

outside the collocated protected area,
for a reactor licensed under parts 50 or
53 of this chapter).
*
*
*
*
*
§ 72.184

[Amended]

141. In § 72.184, in paragraph (a)
remove the phrase ‘‘under part 50 of this
chapter’’ and add in its place the phrase
‘‘under parts 50 or 53 of this chapter’’.
■ 142. Revise § 72.210 to read as
follows:
■

§ 72.210

General license issued.

A general license is hereby issued for
the storage of spent fuel in an
independent spent fuel storage
installation at power reactor sites to
persons authorized to possess or operate
nuclear power reactors under 10 CFR
parts 50, 52, or 53.
■ 143. In § 72.212, revise paragraph
(b)(8) to read as follows:
§ 72.212 Conditions of general license
issued under § 72.210.

*

*
*
*
*
(b) * * *
(8) Before use of the general license,
determine whether activities related to
storage of spent fuel under this general
license involve a change in the facility
Technical Specifications or require a
license amendment for the facility
pursuant to §§ 50.59(c) or 53.1550 of
this chapter. Results of this
determination must be documented in
the evaluations made in paragraph (b)(5)
of this section.
*
*
*
*
*
■ 144. In § 72.218, revise paragraphs (a)
and (b) to read as follows:
§ 72.218

Termination of licenses.

(a) The notification regarding the
program for the management of spent
fuel at the reactor required by
§§ 50.54(bb) or 53.1060 of this chapter
must include a plan for removal of the
spent fuel stored under this general
license from the reactor site. The plan
must show how the spent fuel will be
managed before starting to
decommission systems and components
needed for moving, unloading, and
shipping this spent fuel.
(b) An application for termination of
a reactor operating license issued under
10 CFR part 50 and submitted under
§ 50.82 of this chapter, or a combined
license issued under 10 CFR part 52 and
submitted under § 52.110 of this
chapter, or a reactor operating or
combined license under 10 CFR part 53
and submitted under § 53.1070 of this
chapter must contain a description of
how the spent fuel stored under this

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Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules

general license will be removed from
the reactor site.
*
*
*
*
*
PART 73—PHYSICAL PROTECTION OF
PLANTS AND MATERIALS
145. The authority citation for 10 CFR
part 73 continues to read as follows:

■

Authority: Atomic Energy Act of 1954,
secs. 53, 147, 149, 161, 161A, 170D, 170E,
170H, 170I, 223, 229, 234, 1701 (42 U.S.C.
2073, 2167, 2169, 2201, 2201a, 2210d, 2210e,
2210h, 2210i, 2273, 2278a, 2282, 2297f);
Energy Reorganization Act of 1974, secs. 201,
202 (42 U.S.C. 5841, 5842); Nuclear Waste
Policy Act of 1982, secs. 135, 141 (42 U.S.C.
10155, 10161); 44 U.S.C. 3504 note.
Section 73.37(b)(2) also issued under sec.
301, Pub. L. 96–295, 94 Stat. 789 (42 U.S.C.
5841 note).

146. In § 73.1, revise paragraph
(b)(1)(i) to read as follows:

■

§ 73.1

[Amended]

*

*
*
*
*
(b) * * *
(1) * * *
(i) The physical protection of
production and utilization facilities
licensed under parts 50, 52, or 53 of this
chapter,
*
*
*
*
*
■ 147. In § 73.2, revise the introductory
text and paragraph (a) to read as follows:
§ 73.2

Definitions.

As used in this part:
(a) Terms defined in parts 50, 52, 53,
70, and 95 of this chapter have the same
meaning when used in this part.
*
*
*
*
*
■ 148. In § 73.8, revise paragraph (b) to
read as follows:
§ 73.8 Information collection
requirements: OMB approval.

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*

*
*
*
*
(b) The approved information
collection requirements contained in
this part appear in §§ 73.5, 73.15, 73.17,
73.20, 73.21, 73.24, 73.25, 73.26, 73.27,
73.37, 73.40, 73.45, 73.46, 73.50, 73.54,
73.55, 73.56, 73.57, 73.58, 73.60, 73.67,
73.70, 73.72, 73.73, 73.74, 73.77, 73.100,
73.110, 73.120, 73.1200, 73.1205,
73.1210, 73.1215, and appendices B and
C to this part.
*
*
*
*
*
■ 149. In § 73.50, revise the introductory
text to read as follows:
§ 73.50 Requirements for physical
protection of licensed activities.

Each licensee who is not subject to
§ 73.51, but who possesses, uses, or
stores formula quantities of strategic
special nuclear material that are not
readily separable from other radioactive

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material and which have a total external
radiation level in excess of 1 gray (100
rad) per hour at a distance of 1 meter
(3.3 feet) from any accessible surfaces
without intervening shielding other
than at a nuclear reactor facility
licensed under parts 50, 52, or 53 of this
chapter, shall comply with the
following:
*
*
*
*
*
■ 150. In § 73.55, revise paragraphs
(a)(4) and (6), (i)(4)(iii), (l)(1), (l)(7)(ii),
(p)(1)(i), (r)(2) and (r)(4)(iii) to read as
follows:
§ 73.55 Requirements for physical
protection of licensed activities in nuclear
power reactors against radiological
sabotage.

(a) * * *
(4) Applicants for an operating license
under the provisions of part 50 or part
53 of this chapter or holders of a
combined license under the provisions
of part 52 or part 53 of this chapter shall
implement the requirements of this
section before fuel is allowed onsite
(protected area).
*
*
*
*
*
(6) Applicants for an operating license
under the provisions of part 50 or part
53 of this chapter, or holders of a
combined license under the provisions
of part 52 or part 53 of this chapter that
do not reference a standard design
certification or reference a standard
design certification issued after May 26,
2009, shall meet the requirement of
§ 73.55(i)(4)(iii).
*
*
*
*
*
(i) * * *
(4) * * *
(iii) Applicants for an operating
license under the provisions of part 50
of this chapter, or holders of a combined
license under the provisions of part 52
of this chapter, or licensees under part
53 of this chapter that elect to
demonstrate compliance with § 73.55,
consistent with § 53.860(a)(2) of this
chapter, shall construct, locate, protect,
and equip both the central and
secondary alarm stations to the
standards for the central alarm station
contained in this section. Both alarm
stations shall be equal and redundant,
such that all functions needed to satisfy
the requirements of this section can be
performed in both alarm stations.
*
*
*
*
*
(l) * * *
(1) Commercial nuclear power
reactors licensed under 10 CFR parts 50,
52, or 53 and authorized to use special
nuclear material in the form of MOX
fuel assemblies containing up to 20
weight percent PuO2 shall, in addition
to demonstrating compliance with the

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requirements of this section, protect unirradiated MOX fuel assemblies against
theft or diversion as described in this
paragraph.
*
*
*
*
*
(7) * * *
(ii) Additional measures for the
physical protection of un-irradiated
MOX fuel assemblies containing greater
than 20 weight percent PuO2 shall be
determined by the Commission on a
case-by-case basis and documented
through license amendment in
accordance with §§ 50.90 or 53.1510 of
this chapter.
*
*
*
*
*
(p) * * *
(1) * * *
(i) Under §§ 50.54(x) and (y) or
53.740(h) of this chapter, the licensee
may suspend any security measures
under this section in an emergency
when this action is immediately needed
to protect the public health and safety
and no action consistent with license
conditions and technical specifications
that can provide adequate or equivalent
protection is immediately apparent.
This suspension of security measures
must be approved as a minimum by a
licensed senior operator before taking
this action.
*
*
*
*
*
(r) * * *
(2) The licensee shall submit
proposed alternative measure(s) to the
Commission for review and approval
under §§ 50.4 and 50.90, or §§ 53.040
and 53.1510 of this chapter before
implementation.
*
*
*
*
*
(4) * * *
(iii) Based on comparison of the costs
of the alternative measures to the costs
of demonstrating compliance with the
Commission’s requirements using the
essential elements of §§ 50.109 or
53.1590 of this chapter, the costs of fully
demonstrating compliance with the
Commission’s requirements are not
justified by the protection that would be
provided.
■ 151. In § 73.56, revise paragraph (a)(3)
to read as follows:
§ 73.56 Personnel access authorization
requirements for nuclear power plants.

(a) * * *
(3) Each applicant for an operating
license under the provisions of part 50
of this chapter, each holder of a
combined license under the provisions
of part 52 of this chapter, and applicants
for an operating license or holders of a
combined license under part 53 of this
chapter that do not meet the
requirements of § 53.860(a)(2) of this
chapter, shall implement the

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requirements of this section before fuel
is allowed on site (protected area).
*
*
*
*
*
■ 152. In § 73.57, revise paragraph (a)(3)
to read as follows:
§ 73.57 Requirements for criminal history
records checks of individuals granted
unescorted access to a nuclear power
facility, a non-power reactor, or access to
Safeguards Information.

(a) * * *
(3) Before receiving its operating
license under 10 CFR parts 50 or 53 or
before the Commission makes its
finding under §§ 52.103(g) or 53.1452(g)
of this chapter, each applicant for a
license to operate a nuclear power
reactor (including an applicant for a
combined license) or a non-power
reactor may submit fingerprints for
those individuals who will require
unescorted access to the nuclear power
facility or non-power reactor facility.
*
*
*
*
*
■ 153. In § 73.58, revise paragraph (a) to
read as follows:
§ 73.58 Safety/security interface
requirements for nuclear power reactors.

(a) Each operating nuclear power
reactor licensee with a license issued
under parts 50, 52, or 53 of this chapter
shall comply with the requirements of
this section.
*
*
*
*
*
■ 154. In § 73.67, revise paragraphs (d)
introductory text and (f) introductory
text to read as follows:
§ 73.67 Licensee fixed site and in-transit
requirements for the physical protection of
special nuclear material of moderate and
low strategic significance.

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*

*
*
*
*
(d) Fixed site requirements for special
nuclear material of moderate strategic
significance. Each licensee who
possesses, stores, or uses quantities and
types of special nuclear material of
moderate strategic significance at a fixed
site or contiguous sites, except as
allowed by paragraph (b)(2) of this
section and except those who are
licensed to operate a nuclear power
reactor pursuant to part 50 or part 53,
provided that the special nuclear
material is located within a protected
area and protected under § 73.55 or
§ 73.100, shall:
*
*
*
*
*
(f) Fixed site requirements for special
nuclear material of low strategic
significance. Each licensee who
possesses, stores, or uses special nuclear
material of low strategic significance at
a fixed site or contiguous sites, except
those who are licensed to operate a
nuclear power reactor pursuant to part

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50 or part 53, provided that the special
nuclear material is located within a
protected area and protected under
§ 73.55 or § 73.100, shall:
*
*
*
*
*
■ 155. In § 73.77, revise paragraphs (a),
(b)(1), (c)(6) and (7) to read as follows:
§ 73.77

Cybersecurity event notifications.

(a) Each licensee subject to the
provisions of §§ 73.54 or 73.110 shall
notify the NRC Headquarters Operations
Center via the Emergency Notification
System (ENS), under paragraph (c) of
this section:
(1) Within one hour after discovery of
a cyberattack that adversely impacted:
(i) Safety-related or important-tosafety functions, security functions, or
emergency preparedness functions
(including offsite communications); or
that compromised support systems and
equipment resulting in adverse impacts
to safety, security, or emergency
preparedness functions within the scope
of § 73.54; or,
(ii) Functions performed by digital
assets that would prevent a postulated
fission product release resulting in
offsite doses exceeding the values in
§ 53.210 of this chapter, or functions
performed by digital assets used by the
licensee for implementing the physical
security requirements in § 53.860(a) of
this chapter.
(2) Within 4 hours:
(i) After discovery of a cyberattack
that could have caused an adverse
impact to:
(A) Safety-related or important-tosafety functions, security functions, or
emergency preparedness functions
(including offsite communications); or
that could have compromised support
systems and equipment, which if
compromised, could have adversely
impacted safety, security, or emergency
preparedness functions within the scope
of § 73.54; or,
(B) Functions performed by digital
assets that would prevent a postulated
fission product release resulting in
offsite doses exceeding the values in
§ 53.210 of this chapter, or functions
performed by digital assets used by the
licensee for implementing the physical
security requirements in § 53.860(a) of
this chapter.
(ii) After discovery of a suspected or
actual cyberattack initiated by personnel
with physical or electronic access to
digital computer and communication
systems and networks within the scope
of §§ 73.54 or 73.110.
(iii) After notification of a local, State,
or other Federal agency (e.g., law
enforcement, Federal Bureau of
Investigation (FBI), etc.) of an event
related to the licensee’s implementation

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87119

of their cybersecurity program for digital
computer and communication systems
and networks within the scope of
§§ 73.54 or 73.110 that does not
otherwise require a notification under
paragraph (a) of this section.
(3) Within 8 hours after receipt or
collection of information regarding
observed behavior, activities, or
statements that may indicate
intelligence gathering or pre-operational
planning related to a cyberattack against
digital computer and communication
systems and networks within the scope
of §§ 73.54 or 73.110.
(b) Twenty-four hour recordable
events. (1) The licensee shall use the site
corrective action program to record
vulnerabilities, weaknesses, failures and
deficiencies in their § 73.54 or § 73.110
cybersecurity program within 24 hours
of their discovery.
*
*
*
*
*
(c) * * *
(6) Declaration of emergencies.
Notifications made to the NRC for the
declaration of an emergency class shall
be performed in accordance with
§§ 50.72 or 53.1630 of this chapter, as
applicable.
(7) Elimination of duplication.
Separate notifications and reports are
not required for events that are also
reportable under §§ 50.72 and 50.73 or
§§ 53.1630 and 53.1640 of this chapter.
However, these notifications should also
indicate the applicable § 73.77 reporting
criteria.
*
*
*
*
*
■ 156. Add Subpart J consisting of
§§ 73.100 through 73.120 to read as
follows:
Subpart J—Security Requirements at
Commercial Nuclear Plants
Sec.
73.100 Technology-inclusive requirements
for physical protection of licensed
activities at commercial nuclear plants
against radiological sabotage.
73.110 Technology-inclusive requirements
for protection of digital computer and
communication systems and networks.
73.120 Access authorization program for
commercial nuclear plants.

Subpart J—Security Requirements at
Commercial Nuclear Plants
§ 73.100 Technology-inclusive
requirements for physical protection of
licensed activities at commercial nuclear
plants against radiological sabotage.

(a) Introduction. (1) Each licensee that
is licensed to operate a commercial
nuclear plant under 10 CFR part 53 and
elects to implement the requirements of
this section must do so through its
physical security plan, training and
qualification plan, safeguards

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contingency plan, and cybersecurity
plan, referred to collectively hereafter as
‘‘security plans,’’ before initial fuel load
into the reactor (or, for a fueled
manufactured reactor, before initiating
the physical removal of any one of the
independent physical mechanisms to
prevent criticality required under
§ 53.620(d)(1) of this chapter).
(2) The security plans must identify,
describe, and account for site-specific
conditions that affect the licensee’s
capability to satisfy the requirements of
this section.
(b) General performance objective and
requirements. (1) The licensee must
establish, implement, and maintain a
physical protection program and a
security organization, which will have
as their objective to provide reasonable
assurance that activities involving
special nuclear material are not inimical
to the common defense and security and
do not constitute an unreasonable risk
to the public health and safety.
(2) To satisfy the general performance
objective of paragraph (b)(1) of this
section, the physical protection program
must protect against the design basis
threat of radiological sabotage as stated
in § 73.1. Specifically, the licensee
must—
(i) Ensure that the physical protection
program capabilities to protect against
the design basis threat of radiological
sabotage are maintained at all times; and
(ii) Provide defense in depth in
achieving performance requirements
through the integration of engineered
systems, administrative controls, and
management measures.
(3) The physical protection program
must be designed and implemented to
achieve and maintain the reliability and
availability of structures, systems, and
components (SSCs) required for
demonstrating compliance with the
following performance requirements at
all times:
(i) Intrusion detection. The licensee
must be capable of detecting attempted
and actual unauthorized access to
interior and exterior areas containing
SSCs needed to implement safety and
security functions.
(ii) Intrusion assessment. The licensee
must be capable of timely assessment
for determining the cause of a detected
intrusion.
(iii) Security communication. The
licensee must be capable of continuous
security communications.
Communication systems must account
for design basis threats that can
interrupt or interfere with continuity or
integrity of communications.
(iv) Security response. The physical
protection program must be designed to
provide timely security response to

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interdict and neutralize adversary
attacks up to and including the design
basis threat of radiological sabotage. The
physical protection program must be
designed to provide layers of security
response, with each layer assuring that
a single failure does not result in the
loss of capability to neutralize the
design basis threat adversary.
Structures, systems, and components
relied on for delay functions must be
designed to allow for timely security
responses to adversary attacks with
adequate defense in depth.
(A) The security response may rely on
the use of onsite responders, law
enforcement or other offsite armed
responders, or a combination thereof, to
fulfill the interdiction and
neutralization functions required by
paragraph (b)(3)(iv) of this section. A
licensee relying entirely or partially on
law enforcement or other offsite armed
responders must—
(1) Maintain the capability to detect,
assess, interdict, and neutralize threats
as required by paragraphs (b)(3)(i),
(b)(3)(ii), and (b)(3)(iv) of this section;
(2) Provide adequate delay to enable
law enforcement or other offsite armed
responders to fulfill the interdiction and
neutralization functions for threats up to
and including the design basis threat of
radiological sabotage;
(3) Provide necessary information
about the facility and make available
periodic training to law enforcement or
other offsite armed responders who will
fulfill the interdiction and
neutralization functions for threats up to
and including the design basis threat of
radiological sabotage;
(4) Fully describe in the safeguards
contingency plan the role that law
enforcement or other offsite armed
responders will play in the licensee’s
protective strategy. The description
must provide sufficient detail to enable
the NRC to determine that the licensee’s
physical protection program provides
reasonable assurance of adequate
protection against threats up to and
including the design basis threat of
radiological sabotage; and
(5) Identify criteria and measures to
compensate for the degradation or
absence of law enforcement or other
offsite armed responders and propose
suitable compensatory measures that
meet the requirements of paragraph
(h)(3) of this section to address this
degradation.
(B) For licensees relying entirely or
partially on law enforcement responders
to fulfill the interdiction and
neutralization functions required by
paragraph (b)(3)(iv) of this section, the
training and qualification requirements
related to armed response personnel in

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paragraphs (c) and (e) of this section do
not apply to law enforcement
responders. The licensee shall continue
to satisfy the performance evaluation
requirements in paragraph (g) of this
section for all armed response
personnel, including law enforcement.
(v) Protecting against land and
waterborne vehicle bomb assaults. The
licensee must be capable of protecting
the plant against the design basis threat
vehicle bomb assault. The methods that
are relied on to protect against a design
basis threat land vehicle and waterborne
vehicle bomb assault must be designed
to protect the reactor building and
structures containing safety- or securityrelated systems, and components from
explosive effects.
(vi) Access control portals. The
licensee must be capable of detecting
and denying unauthorized access to
persons and pass-through of contraband
materials (e.g., weapons, incendiaries,
explosives) to protected areas.
(4) The licensee must meet the
requirements related to target sets in
§ 73.55(f).
(5) The licensee must identify and
analyze site-specific conditions,
including target sets, that may affect the
physical protection program needed to
implement the requirements of this
section. The licensee must account for
these conditions in demonstrating
compliance with the requirements of
this section.
(6) The licensee must establish,
implement, and maintain a performance
evaluation program to assess the
effectiveness of the licensee’s
implementation of the physical
protection program to protect against
the design basis threat of radiological
sabotage.
(7) The licensee must establish,
implement, and maintain an access
authorization program under § 73.56
and must describe the program in the
physical security plan.
(8) The licensee must establish,
implement, and maintain a
cybersecurity program under §§ 73.54 or
73.110 and must describe the program
in the cybersecurity plan.
(9) The licensee must establish,
implement, and maintain an insider
mitigation program and must describe
the program in the physical security
plan.
(i) The insider mitigation program
must monitor the initial and continuing
trustworthiness and reliability of
individuals granted or retaining
unescorted access or unescorted access
authorization to a protected or vital
area, and implement defense-in-depth
methodologies to minimize the potential
for an insider (active, passive, or both)

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to adversely affect, either directly or
indirectly, the licensee’s capability to
protect against radiological sabotage.
(ii) The insider mitigation program
must integrate elements of—
(A) The access authorization program
under § 73.56;
(B) The fitness-for-duty program
under 10 CFR part 26;
(C) The cybersecurity program under
§§ 73.54 or 73.110; and
(D) The physical protection program
under this section.
(10) The licensee must have the
capability to track, trend, correct, and
prevent recurrence of failures and
deficiencies in the implementation of
the requirements of this section.
(11) Implementation of security plans
and associated procedures must be
coordinated with other onsite plans and
procedures to preclude conflict during
both normal and emergency conditions
and ensure the adequate management of
the safety and security interface.
(12)(i) The licensee must ensure that
the firearms background check
requirements of § 73.17 of this part are
met for all members of the security
organization whose official duties
require access to covered weapons or
who inventory enhanced weapons.
(ii) The provisions of this paragraph
are only applicable to licensees subject
to this section that are also subject to the
firearms background check provisions of
§ 73.17 of this part.
(c) Security organization. The licensee
must establish and maintain a security
organization that is staffed, trained,
qualified, and equipped to implement
the physical protection program under
the requirements of this section.
(1) The licensee must establish a
management system for maintaining and
implementing security policies and
procedures to implement the
requirements of this section and the
security plans.
(2) Implementing procedures must
document the conduct of security
operations, security design and
configuration controls, maintenance,
training and qualification, and
contingency responses.
(3) The licensee must—
(i) Establish a process for the approval
of designs, policies, processes, and
procedures and changes by the
individual with overall responsibility
for the physical protection program; and
(ii) Ensure that revisions and changes
to the physical protection program and
implementing policies, processes, and
procedures satisfy the requirements of
this section.
(4) The licensee must retain, in
accordance with § 73.70, all analyses,
assessments, calculations, and

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descriptions of the technical basis for
demonstrating compliance with the
performance requirements of
§ 73.100(b). The licensee must protect
these records in accordance with the
requirements for protecting safeguards
information in §§ 73.21 and 73.22.
(5) The licensee may not permit any
individual to implement any part of the
physical protection program unless the
individual has been trained, equipped,
and qualified to perform their assigned
duties and responsibilities in
accordance with the training and
qualification plan.
(d) Search requirements. The licensee
must establish and implement searches
of individuals, vehicles, and materials
to detect and prevent the introduction
into the protected area of firearms,
explosives, incendiary devices, or other
items and material which could be used
to commit radiological sabotage.
(e) Training and qualification
program. The licensee must establish
and maintain a training and
qualification program that ensures
personnel who are responsible for the
physical protection of the facility
against radiological sabotage are able to
effectively perform their assigned
security-related job duties for
implementing the requirements of this
section and must describe the program
in the training and qualification plan.
(f) Security reviews. The licensee must
establish and implement security
reviews to assess the effectiveness of the
implementation of the physical
protection program. Security reviews
must be performed by individuals
independent of those personnel
responsible for program management
and any individual who has direct
responsibility for implementing the
onsite physical protection program.
(1) The licensee must review each
element of the physical protection
program at a frequency commensurate
with the importance or significance to
safety of plant operations to ensure
timely identification and documentation
of vulnerabilities, improvements, and
corrective actions. The objective of these
reviews must be maintaining effective
implementation of the engineered and
administrative controls required to
achieve the physical protection program
functions and the management system
required to implement programs and
requirements in this section.
(2) The licensee must establish and
perform self-assessments to ensure the
effective implementation of the physical
protection program functions of
detection, assessment, communication,
delay, and interdiction and
neutralization to protect against the
design basis threat of radiological

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sabotage. The licensee must perform
design verification and assessments of
the capabilities of active and passive
engineering systems relied on to protect
against the design basis threat.
(3) Reviews of the security program
must include, but are not limited to, an
audit of the effectiveness of the physical
protection program, security plans,
implementing procedures, cybersecurity
programs, safety/security interface
activities, the testing, maintenance, and
calibration program, and response
commitments by local, State, and
Federal law enforcement authorities.
(4) The results and recommendations
of the onsite physical protection
program reviews, management’s
findings regarding program
effectiveness, and any actions taken as
a result of recommendations from prior
program reviews, must be documented
in a report and must be maintained in
an auditable form and available for
inspection.
(g) Performance evaluation. Licensee
performance evaluations must include
methods appropriate and necessary to
assess, test, and challenge the
integration of the physical protection
program’s functions to protect against
the design basis threat, including
measures to protect against cyberattack
and engineered systems designed to
protect against the design basis threat
standalone ground vehicle bomb attack.
(1) The licensee must establish the
frequencies for performance evaluations
commensurate with the security
significance of the physical protection
program.
(2) The licensee must document
processes and procedures for
implementing the performance
evaluations. The licensee must maintain
records, including results, findings, and
corrective actions identified during the
performance evaluations.
(h) Maintenance, testing, and
calibration and corrective actions. (1)
The licensee must ensure that security
SSCs, including supporting systems, are
inspected, tested, and calibrated for
operability and performance at intervals
necessary and sufficient to meet the
requirements of this section.
(2) The licensee must implement
corrective actions to ensure resolution
of identified vulnerabilities and
deficiencies to meet the requirements of
this section.
(3) The licensee must establish and
implement timely compensatory
measures for degraded or inoperable
security SSCs to meet the requirements
of this section. Compensatory measures
must provide a level of protection that
is equivalent to the protection that was
provided prior to the degradation or

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inoperability of the security structures,
systems, or components.
(4) The licensee must document
processes and procedures and maintain
records for implementing the corrective
actions, compensatory measures, and
maintenance, inspection, testing, and
calibration of security SSCs.
(i) Suspension of security measures.
(1) The licensee may suspend
implementation of affected
requirements of this section in
accordance with § 53.740(h) of this
chapter under the following conditions:
(i) In an emergency, when action is
immediately needed to protect the
public health and safety; and
(ii) During severe weather, when the
suspension of affected security
measures is immediately needed to
protect the personal health and safety of
personnel.
(2) Suspended security measures must
be reinstated as soon as conditions
permit.
(3) The suspension of security
measures must be reported and
documented in accordance with the
provisions of §§ 73.1200 and 73.1205.
(j) Records. (1) The Commission may
inspect, copy, retain, and remove all
reports, records, and documents
required to be kept by Commission
regulations, orders, or license
conditions, whether the reports, records,
and documents are kept by the licensee
or a contractor.
(2) The licensee must maintain all
records required to be kept by
Commission regulations, orders, or
license conditions, until the
Commission terminates the license for
which the records were developed and
must maintain superseded portions of
these records for at least 3 years after the
record is superseded, unless otherwise
specified by the Commission.
(3) If a contracted security force is
used to implement the onsite physical
protection program, the licensee’s
written agreement with the contractor
must be retained by the licensee as a
record for the duration of the contract.
(4) Review and audit reports must be
available for inspection, for a period of
3 years.

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§ 73.110 Technology-inclusive
requirements for protection of digital
computer and communication systems and
networks.

(a) Each licensee that is licensed to
operate a commercial nuclear plant
under 10 CFR part 53 and elects to
implement the requirements of this
section must establish, implement, and
maintain a cybersecurity program that is
commensurate with the potential
consequences resulting from

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cyberattacks, up to and including the
design basis threat as described in
§ 73.1. The cybersecurity program must
provide reasonable assurance that
digital computer and communication
systems and networks are adequately
protected against cyberattacks that are
capable of causing the following
consequences:
(1) Adversely impacting the functions
performed by digital assets that would
prevent a postulated fission product
release resulting in offsite doses
exceeding the values in § 53.210 of this
chapter.
(2) Adversely impacting the functions
performed by digital assets used by the
licensee for implementing the physical
security requirements in § 53.860(a) of
this chapter.
(b) To protect digital computer and
communication systems and networks
associated with the functions described
in paragraphs (a)(1) and (2), the licensee
must—
(1) Analyze the potential
consequences resulting from
cyberattacks on digital computer and
communication systems and networks
and identify those assets that must be
protected to demonstrate compliance
with paragraph (a) of this section; and
(2) Implement the cybersecurity
program in accordance with paragraph
(d) of this section.
(c) The licensee must comply with the
requirements in § 73.54(a)(2) for the
systems and networks identified in
paragraph (b)(1) of this section in a
manner that is commensurate with the
potential consequences resulting from
cyberattacks.
(d) The cybersecurity program must
be designed in a manner that is
commensurate with the potential
consequences resulting from
cyberattacks through the following
steps:
(1) Implement security controls to
protect the assets identified under
paragraph (b)(1) of this section from
cyberattacks, commensurate with their
safety and security significance;
(2) Apply and maintain defense-indepth protective strategies to ensure the
capability to detect, delay, respond to,
and recover from cyberattacks capable
of causing the consequences identified
in paragraph (a) of this section;
(3) Mitigate the adverse effects of
cyberattacks capable of causing the
consequences identified in paragraph (a)
of this section; and
(4) Ensure that the functions of
protected assets identified under
paragraph (b)(1) of this section are not
adversely impacted due to cyberattacks.
(e) The licensee must implement the
following requirements in a manner that

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is commensurate with the potential
consequences resulting from
cyberattacks:
(1) As part of the cybersecurity
program, the licensee must comply with
the requirements in § 73.54(d)(1), (2),
and (4), and must ensure that
modifications to assets, identified under
paragraph (b)(1) of this section are
evaluated before implementation to
ensure that the cybersecurity
performance objectives identified in
paragraph (a) of this section are
maintained.
(2) The licensee must establish,
implement, and maintain a
cybersecurity plan that implements the
cybersecurity program requirements of
this section.
(i) The cybersecurity plan must
describe how the requirements of this
section will be implemented and must
account for the site-specific conditions
that affect implementation.
(ii) The cybersecurity plan must
include measures for incident response
and recovery for cyberattacks. The
cybersecurity plan must include the
analysis identified under paragraph
(b)(1) of this section and describe how
the licensee will—
(A) Apply and maintain defense-indepth protective strategies as required
in paragraph (d)(2) of this section;
(B) Maintain the capability for timely
detection and response to cyberattacks;
(C) Mitigate the consequences of
cyberattacks;
(D) Correct exploited vulnerabilities;
and
(E) Restore affected systems,
networks, and/or equipment affected by
cyberattacks.
(3) The licensee must develop and
maintain written policies and
implementing procedures to implement
the cybersecurity plan. Policies,
implementing procedures, and other
supporting technical information used
by the licensee need not be submitted
for Commission review and approval as
part of the cybersecurity plan but are
subject to inspection by NRC staff on a
periodic basis.
(4) The licensee must establish and
implement cybersecurity reviews to
assess the effectiveness of the
implementation of the cybersecurity
program.
(i) The licensee must review each
element of the cybersecurity program at
a frequency commensurate with the
importance or significance to safety of
plant operations to ensure timely
identification and documentation of
vulnerabilities, improvements, and
corrective actions.
(ii) Cybersecurity reviews must be
performed by individuals independent

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of those personnel responsible for
program management and any
individual who has direct responsibility
for implementing the cybersecurity
program.
(iii) The licensee must establish and
perform self-assessments to ensure the
effective implementation of the
cybersecurity program.
(iv) The results and recommendations
of the cybersecurity program reviews,
management’s findings regarding
program effectiveness, and any actions
taken as a result of recommendations
from prior program reviews, must be
documented in a report and must be
maintained in an auditable form and
available for inspection.
(5) The licensee must retain all
records and supporting technical
documentation required to demonstrate
compliance with the requirements of
this section as a record until the
Commission terminates the license for
which the records were developed and
must maintain superseded portions of
these records for at least three (3) years
after the record is superseded, unless
otherwise specified by the Commission.

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§ 73.120 Access authorization program for
commercial nuclear plants.

(a) Introduction and scope. Each
applicant for an operating license or a
holder of a combined license under 10
CFR part 53 must establish, maintain,
and implement an access authorization
program before initial fuel load into the
reactor (or, for a fueled manufactured
reactor, before initiating the physical
removal of any one of the independent
physical mechanisms to prevent
criticality required under § 53.620(d)(1)
of this chapter). The requirements in
this section apply to licensees satisfying
the criterion in § 53.860(a)(2)(i) of this
chapter.
(b) Applicability. (1) The following
individuals must be subject to an access
authorization program under this
section:
(i) Any individual to whom a licensee
intends to grant unescorted access to a
commercial nuclear plant protected
area, vital area, or controlled access area
where licensed material is used or
stored;
(ii) Any individual whose duties and
responsibilities permit the individual to
take actions by electronic means, either
on site or remotely, that could adversely
impact the licensee’s or applicant’s
operational safety, security, or
emergency preparedness;
(iii) Any individual who has
responsibilities for implementing a
licensee’s or applicant’s protective
strategy, including armed security force
officers, alarm station operators, and

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tactical response team leaders but not
including Federal, State, or local law
enforcement personnel; and
(iv) The licensee or applicant access
authorization program reviewing official
or contractor or vendor access
authorization program reviewers.
(2) The licensee or applicant may
subject other individuals, including
employees of a contractor or a vendor
who are designated in access
authorization program procedures, to an
access authorization program that
demonstrates compliance with the
requirements of this section.
(c) General performance objectives
and requirements. Each licensee’s or
applicant’s access authorization
program under this section must
demonstrate that the individuals who
are specified in paragraph (b) of this
section are trustworthy and reliable,
such that they do not constitute an
unreasonable risk to public health and
safety or the common defense and
security. The licensee’s access
authorization program must maintain
the capabilities for demonstrating
compliance with the following
performance requirements:
(1) Background investigation. (i)(A)
Licensees and applicants must ensure
that any individual seeking initial
unescorted access or to maintain
unescorted access is subject to a
background investigation.
(B) Background investigations must
include the program elements contained
under § 37.25 of this chapter and must
also include a credit history evaluation.
(ii) Background investigations must
include fingerprinting and an FBI
identification and criminal history
records check in accordance with
§ 37.27 of this chapter.
(iii) Licensees must have the informed
and signed consent of the subject
individual to initiate a background
investigation. This consent must
include authorization to share personal
information with other individuals or
organizations as necessary to complete
the background investigation. A signed
consent must be obtained prior to any
reinvestigation. The subject individual
may withdraw his or her consent at any
time. Licensees must inform the
individual that—
(A) If an individual withdraws his or
her consent, the licensee may not
initiate any elements of the background
investigation that were not in progress
at the time the individual withdrew his
or her consent; and
(B) The withdrawal of consent for the
background investigation is sufficient
cause for denial or termination of
unescorted access authorization.

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(2) Behavioral observation. Licensees,
applicants, contractors, and vendors
must ensure the access authorization
program includes provisions that the
individuals specified in paragraph (b) of
this section are subject to behavioral
observation.
(i) Each person subject to behavioral
observation must communicate to the
licensee or applicant observed behaviors
or activities of individuals that may
constitute an unreasonable risk to the
health and safety of the public and
common defense and security.
(ii) Behavioral observation must
include visual observation, in person or
remotely by video, to detect and
promptly report to plant supervision
any concerns arising from behavioral
observation, including, but not limited
to, concerns related to any questionable
behavior patterns or activities of others.
(3) Self-reporting of legal actions.
Licensees or applicants must inform
personnel who are granted and who
maintain unescorted access of their
responsibilities to self-report to plant
supervision legal actions taken by a law
enforcement authority or court of law
against the individual that could result
in incarceration or a court order or that
requires a court appearance, including
but not limited to an arrest, an
indictment, the filing of charges, or a
conviction, but excluding minor civil
actions or misdemeanors such as
parking violations or speeding tickets,
for any individual who has applied for
unescorted access or who maintains
unescorted access.
(4) Unescorted access. Licensees or
applicants must grant unescorted access
only after the licensee has verified an
individual is trustworthy and reliable. A
list of persons currently approved for
unescorted access to a protected area,
vital area, or controlled access area must
be maintained at all times. Unescorted
access determinations must be reviewed
annually by the reviewing official.
Licensees and applicants must complete
an FBI criminal history record check
update for each individual maintaining
unescorted access, within 10 years of
the last review.
(5) Termination of unescorted access.
Licensees and applicants must promptly
terminate unescorted access when this
access is no longer required or a
reviewing official determines an
individual is no longer trustworthy and
reliable in accordance with this section.
(6) Determination basis for access. (i)
The licensee’s or applicant’s reviewing
official must determine whether to
permit, deny, unfavorably terminate,
maintain, or administratively withdraw
an individual’s unescorted access based
on an evaluation of all of the

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information collected to demonstrate
compliance with the requirements of
this section.
(ii) Licensees and applicants must
provide individuals subject to this
section, prior to any final adverse
determination, the right to complete,
correct, and explain information
obtained as a result of the licensee’s
background investigation pursuant to
§ 37.23(g) of this chapter.
(iii) The licensee’s or applicant’s
reviewing officials are the only
individuals authorized to make
unescorted access determination
decisions. Each licensee or applicant
must name one or more individuals to
be reviewing officials pursuant to the
requirements of § 37.23(b)(2) of this
chapter.
(7) Review procedures. Review
procedures must be established in
accordance with § 37.23(f) of this
chapter, to include provisions for the
notification in writing of individuals
who are denied unescorted access or
who are unfavorably terminated.
(8) Protection of information.
Licensees, applicants, contractors, or
vendors must establish and maintain a
system of files and procedures in
accordance with § 37.31 of this chapter,
to ensure personal information is not
disclosed to unauthorized persons.
(9) Access authorization reviews and
corrective action. Licensees and
applicants must develop, implement,
and maintain procedures for conduct of
access authorization reviews and
corrective actions in accordance with
§ 37.33 of this chapter to ensure the
continuing effectiveness of the access
authorization program and to ensure
that the access authorization program
and program elements are in
compliance with the requirements of
this section. Each licensee and applicant
must be responsible for the continuing
effectiveness of the access authorization
program, including access authorization
program elements that are provided by
the contractors or vendors, and the
access authorization programs of any of
the contractors or vendors that are
accepted by the licensee or applicant.
(10) Records. Licensees, applicants,
and contractors or vendors must
document the processes and procedures
for maintaining records used or created
to establish an individual’s
trustworthiness and reliability or to
document access determinations.
Licensees, applicants, and contractor or
vendors must—
(i) Retain documentation regarding
the trustworthiness and reliability of
individual employees for 3 years from
the date the individual no longer
requires unescorted access;

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(ii) Retain a copy of the current access
authorization program procedures as a
record for 3 years after the procedure is
no longer needed. If any portion of the
procedure is superseded, retain the
superseded material for 3 years after the
record is superseded; and
(iii) Retain the list of persons
approved for unescorted access for 3
years after the list is superseded or
replaced. Records maintained in any
database(s) must be available for NRC
review.
■ 157. In § 73.1200, revise paragraphs
(a) introductory text, (c)(1) introductory
text, (e)(1) introductory text, (e)(3) and
(4), (g)(1) introductory text, (o)(5)(i) and
(o)(6)(i), (r) and (s) to read as follows:
§ 73.1200
events.

Notification of physical security

(a) 15-minute notifications—facilities.
Each licensee subject to the provisions
of § 73.20, § 73.45, § 73.46, § 73.51,
§ 73.55, or § 73.100 must notify the NRC
Headquarters Operations Center, as soon
as possible but within 15 minutes
after—
*
*
*
*
*
(c) * * *
(1) Each licensee subject to the
provisions of §§ 73.20, 73.45, 73.46,
73.50, 73.51, 73.55, 73.60, 73.67, or
73.100 must notify the NRC
Headquarters Operations Center as soon
as possible but no later than 1 hour after
the time of discovery of the following
significant facility security events
involving—
*
*
*
*
*
(e) * * *
(1) Each licensee subject to the
provisions of §§ 73.20, 73.45, 73.46,
73.50, 73.51, 73.55, 73.60, 73.67, or
73.100 must notify the NRC
Headquarters Operations Center within
4 hours after time of discovery of the
following facility security events
involving—
*
*
*
*
*
(3)(i) An event involving a law
enforcement response to the facility that
could reasonably be expected to result
in public or media inquiries and that
does not otherwise require a notification
under paragraphs (a) through (h) of this
section, or in other NRC regulations
such as § 50.72(b)(2)(xi) or
§ 53.1630(b)(2)(v) of this chapter.
(ii) As an exemption, licensees need
not report law enforcement responses to
minor incidents, such as traffic
accidents.
(4) For licensees subject to the
provisions of § 73.55 or § 73.100 of this
part, an event involving the licensee’s
suspension of security measures.
*
*
*
*
*

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(g) * * *
(1) Each licensee subject to the
provisions of § 73.20, § 73.45, § 73.46,
§ 73.50, § 73.51, § 73.55, § 73.60, § 73.67,
or § 73.100 must notify the NRC
Headquarters Operations Center within
8 hours after time of discovery of the
following facility security program
failures involving—
*
*
*
*
*
(o) * * *
(5) * * *
(i) Licensees must establish the
requested continuous communications
channel once the licensee has
completed other required notifications
under this section, § 50.72 of this
chapter, appendix E to part 50 of this
chapter, § 53.1630 of this chapter,
§ 70.50 of this chapter; or § 72.75 of this
chapter; as appropriate.
*
*
*
*
*
(6) * * *
(i) Licensees must establish the
requested continuous communications
channel once the licensee or the
movement control center has completed
other required notifications under this
section, § 50.72 of this chapter,
appendix E to part 50 of this chapter,
§ 53.1630 of this chapter, § 70.50 of this
chapter; § 72.75 of this chapter; or
requested assistance from the LLEA, as
appropriate.
*
*
*
*
*
(r) Declaration of emergencies.
Licensees notifying the NRC of the
declaration of an emergency class must
do so in accordance with §§ 50.72,
53.1630, 63.73, 70.50, and 72.75 of this
chapter, as applicable.
(s) Elimination of duplication.
Licensees with notification obligations
under paragraphs (a) through (h), (m),
and (n) of this section and §§ 50.72,
53.1630, 63.73, 70.50, and 72.75 of this
chapter may notify the NRC of events in
a single communication. This
communication must identify each
regulation under which the licensee is
reporting.
*
*
*
*
*
■ 158. In § 73.1205, revise paragraph
(b)(2) to read as follows:
§ 73.1205 Written follow-up reports of
physical security events.

*

*
*
*
*
(b) * * *
(2)(i) Licensees subject to § 50.73 or
§ 53.1640 of this chapter must prepare
the written follow-up report on NRC
Form 366.
(ii) Licensees not subject to § 50.73 or
§ 53.1640 of this chapter must prepare
the written follow-up report in a letter
format.
*
*
*
*
*

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Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules
■

159. In § 73.1210, revise paragraphs
(a)(1) and (b)(3)(i) to read as follows:

§ 74.31 Nuclear material control and
accounting for special nuclear material of
low strategic significance.

§ 73.1210 Recordkeeping of physical
security events.

(a) General performance objectives.
Each licensee who is authorized to
possess and use more than one effective
kilogram of special nuclear material of
low strategic significance, excluding
sealed sources, at any site or contiguous
sites subject to control by the licensee,
other than a production or utilization
facility licensed pursuant to parts 50,
53, or 70 of this chapter, or operations
involved in waste disposal, shall
implement and maintain a Commission
approved material control and
accounting system that will achieve the
following objectives:
*
*
*
*
*
■ 164. In § 74.41, revise paragraph (a)
introductory text to read as follows:

(a) * * *
(1) Licensees with facilities or
shipment activities subject to the
provisions of § 73.20, § 73.25, § 73.26,
§ 73.27, § 73.37, § 73.45, § 73.46, § 73.50,
§ 73.51, § 73.55, § 73.60, § 73.67, or
§ 73.100, must record the physical
security events and conditions adverse
to security that are specified in
paragraphs (c) through (f) of this section.
*
*
*
*
*
(b) * * *
(3)(i) Licensees must record these
physical security events and conditions
adverse to security in either a standalone safeguards event log or as part of
the licensee’s corrective action program,
as specified under the applicable quality
assurance program provisions of parts
50, 52, 53, 60, 63, 70, and 72 of this
chapter, or both.
*
*
*
*
*
■ 160. In § 73.1215, revise paragraph
(d)(1) introductory text to read as
follows:
§ 73.1215

Suspicious activity reports.

*

*
*
*
*
(d) * * *
(1) For licensees subject to the
provisions of §§ 73.20, 73.45, 73.46,
73.50, 73.51, 73.55, 73.60, 73.67, or
73.100, the licensees must report
activities they assess are suspicious.
Examples include, but are not limited
to, the following:
*
*
*
*
*
■ 161. In appendix B to part 73, revise
Definitions introductory text to read as
follows:
Appendix B to Part 73—General
Criteria for Security Personnel
*

*

*

*

*

Definitions
Terms defined in parts 50, 53, 70, and 73
of this chapter have the same meaning when
used in this appendix.

*

*

*

*

*

PART 74—MATERIAL CONTROL AND
ACCOUNTING OF SPECIAL NUCLEAR
MATERIAL
162. The authority citation for 10 CFR
part 74 continues to read as follows:

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■

Authority: Atomic Energy Act of 1954,
secs. 53, 57, 161, 182, 223, 234, 1701 (42
U.S.C. 2073, 2077, 2201, 2232, 2273, 2282,
2297f); Energy Reorganization Act of 1974,
secs. 201, 202 (42 U.S.C. 5841, 5842); 44
U.S.C. 3504 note.

163. In § 74.31, revise paragraph (a)
introductory text to read as follows:

■

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§ 74.41 Nuclear material control and
accounting for special nuclear material of
moderate strategic significance.

(a) General performance objectives.
Each licensee who is authorized to
possess special nuclear material (SNM)
of moderate strategic significance or
SNM in a quantity exceeding one
effective kilogram of strategic special
nuclear material in irradiated fuel
reprocessing operations other than as
sealed sources and to use this material
at any site other than a nuclear reactor
licensed pursuant to parts 50 or 53 of
this chapter; or as reactor irradiated
fuels involved in research,
development, and evaluation programs
in facilities other than irradiated fuel
reprocessing plants; or an operation
involved with waste disposal, shall
establish, implement, and maintain a
Commission-approved material control
and accounting (MC&A) system that will
achieve the following performance
objectives:
*
*
*
*
*
■ 165. In § 74.51, revise paragraph (a)
introductory text to read as follows:
§ 74.51 Nuclear material control and
accounting for strategic special nuclear
material.

(a) General performance objectives.
Each licensee who is authorized to
possess five or more formula kilograms
of strategic special nuclear material
(SSNM) and to use such material at any
site, other than a nuclear reactor
licensed pursuant to parts 50 or 53 of
this chapter, an irradiated fuel
reprocessing plant, an operation
involved with waste disposal, or an
independent spent fuel storage facility
licensed pursuant to part 72 of this
chapter shall establish, implement, and
maintain a Commission-approved
material control and accounting (MC&A)

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87125

system that will achieve the following
objectives:
*
*
*
*
*
PART 75—SAFEGUARDS ON
NUCLEAR MATERIAL—
IMPLEMENTATION OF SAFEGUARDS
AGREEMENTS BETWEEN THE UNITED
STATES AND THE INTERNATIONAL
ATOMIC ENERGY AGENCY
166. The authority citation for 10 CFR
part 75 continues to read as follows:

■

Authority: Atomic Energy Act of 1954,
secs. 53, 63, 103, 104, 122, 161, 223, 234,
1701 (42 U.S.C. 2073, 2093, 2133, 2134, 2152,
2201, 2273, 2282, 2297f); Energy
Reorganization Act of 1974, sec. 201 (42
U.S.C. 5841); Nuclear Waste Policy Act of
1982, secs. 135, 141 (42 U.S.C. 10155, 10161);
44 U.S.C. 3504 note.

167. In § 75.4, revise the introductory
text and paragraph (6) of the definition
for ‘‘Facility’’ to read as follows:

■

§ 75.4

Definitions.

*

*
*
*
*
Unless otherwise defined in this
section, the terms defined in §§ 40.4,
50.2, 53.020, and 70.4 of this chapter
have the same meaning when used in
this part.
*
*
*
*
*
Facility means:
*
*
*
*
*
(6) Any plant or location where the
possession of more than 1 effective
kilogram of nuclear material is licensed
pursuant to 10 CFR part 40, 50, 53, 60,
61, 63, 70, 72, 76, or 150 of this chapter
or an Agreement State license.
*
*
*
*
*
PART 95—FACILITY SECURITY
CLEARANCE AND SAFEGUARDING
OF NATIONAL SECURITY
INFORMATION AND RESTRICTED
DATA
168. The authority citation for 10 CFR
part 95 continues to read as follows:

■

Authority: Atomic Energy Act of 1954,
secs. 145, 161, 223, 234 (42 U.S.C. 2165,
2201, 2273, 2282); Energy Reorganization Act
of 1974, sec. 201 (42 U.S.C. 5841); 44 U.S.C.
3504 note; E.O. 10865, as amended, 25 FR
1583, 3 CFR, 1959–1963 Comp., p. 398; E.O.
12829, 58 FR 3479, 3 CFR, 1993 Comp., p.
570; E.O. 12968, 60 FR 40245, 3 CFR, 1995
Comp., p. 391; E.O. 13526, 75 FR 707, 3 CFR,
2009 Comp., p. 298.

169. In § 95.5, revise the definition for
‘‘License’’ to read as follows:

■

§ 95.5

Definitions.

*

*
*
*
*
License means a license issued under
10 CFR part 50, 52, 53, 54, 60, 63, 70,
or 72.
*
*
*
*
*

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§ 95.39

Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules
[Amended]

170. In § 95.39(a), remove ‘‘part 52’’
and add in its place ‘‘parts 52 or 53.’’

■

PART 140—FINANCIAL PROTECTION
REQUIREMENTS AND INDEMNITY
AGREEMENTS
171. The authority citation for 10 CFR
part 140 continues to read as follows:

■

Authority: Atomic Energy Act of 1954,
secs. 161, 170, 223, 234 (42 U.S.C. 2201,
2210, 2273, 2282); Energy Reorganization Act
of 1974, secs. 201, 202 (42 U.S.C. 5841,
5842); 44 U.S.C. 3504 note.

172. In § 140.2, revise paragraphs
(a)(1) and (2) to read as follows:

■

§ 140.2

Scope.

(a) * * *
(1) To each person who is an
applicant for or holder of a license
issued under 10 CFR part 50, 52, 53, or
54 to operate a nuclear reactor, and
(2) With respect to an extraordinary
nuclear occurrence, to each person who
is an applicant for or holder of a license
to operate a production facility or a
utilization facility (including an
operating license issued under part 50
or part 53 of this chapter and a
combined license under part 52 or part
53 of this chapter), and to other persons
indemnified with respect to the
involved facilities.
*
*
*
*
*
■ 173. Revise § 140.10 to read as
follows:
§ 140.10

Scope.

This subpart applies to each person
who is an applicant for or holder of a
license issued under 10 CFR parts 50, 53
or 54 to operate a nuclear reactor, or is
the applicant for or holder of a
combined license issued under 10 CFR
parts 52, 53, or 54, except licenses held
by persons found by the Commission to
be Federal agencies or nonprofit
educational institutions licensed to
conduct educational activities. This
subpart also applies to persons licensed
to possess and use plutonium in a
plutonium processing and fuel
fabrication plant.
■ 174. In § 140.11, revise paragraph (b)
to read as follows:
§ 140.11 Amounts of financial protection
for certain reactors.
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*

*
*
*
*
(b) In any case where a person is
authorized under 10 CFR parts 50, 52,
53, or 54 to operate two or more nuclear
reactors at the same location, the total
primary financial protection required of
the licensee for all such reactors is the
highest amount which would otherwise
be required for any one of those

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reactors; provided, that such primary
financial protection covers all reactors
at the location.
■ 175. In § 140.12, revise paragraph (c)
to read as follows:

PART 150—EXEMPTIONS AND
CONTINUED REGULATORY
AUTHORITY IN AGREEMENT STATES
AND IN OFFSHORE WATERS UNDER
SECTION 274

§ 140.12 Amount of financial protection
required for other reactors.

■

*

*
*
*
*
(c) In any case where a person is
authorized under 10 CFR parts 50, 52,
53, or 54 to operate two or more nuclear
reactors at the same location, the total
financial protection required of the
licensee for all such reactors is the
highest amount which would otherwise
be required for any one of those
reactors; provided, that such financial
protection covers all reactors at the
location.
*
*
*
*
*
■ 176. Revise § 140.13 to read as
follows:
§ 140.13 Amount of financial protection
required of certain holders of construction
permits and combined licenses under 10
CFR part 52.

Each holder of a 10 CFR part 50 or 10
CFR part 53 construction permit, or a
holder of a combined license under
parts 52 or 53 of this chapter before the
date that the Commission had made the
finding under §§ 52.103(g) or 53.1452(g)
of this chapter, who also holds a license
under part 70 of this chapter authorizing
ownership, possession and storage only
of special nuclear material at the site of
the nuclear reactor for use as fuel in
operation of the nuclear reactor after
issuance of either an operating license
under 10 CFR part 50 or 53, or a
combined license under 10 CFR part 52
or 53, shall, during the period before
issuance of a license authorizing
operation under 10 CFR part 50 or 53,
or the period before the Commission
makes the finding under § 52.103(g) or
§ 53.1452(g) of this chapter, as
applicable, have and maintain financial
protection in the amount of $1,000,000.
Proof of financial protection shall be
filed with the Commission in the
manner specified in § 140.15 before
issuance of the license under part 70 of
this chapter.
■ 177. In § 140.20, revise paragraphs
(a)(1)(i) and (ii) to read as follows:
§ 140.20

Indemnity agreements and liens.

(a) * * *
(1)(i) The effective date of the license
(issued under part 50 or part 53 of this
chapter) authorizing the licensee to
operate the nuclear reactor involved; or
(ii) The date that the Commission
makes the finding under §§ 52.103(g) or
53.1452(g) of this chapter; or
*
*
*
*
*

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178. The authority citation for 10 CFR
part 150 continues to read as follows:

Authority: Atomic Energy Act of 1954,
secs. 11, 53, 81, 83, 84, 122, 161, 181, 223,
234, 274 (42 U.S.C. 2014, 2201, 2231, 2273,
2282, 2021); Energy Reorganization Act of
1974, sec. 201 (42 U.S.C. 5841); Nuclear
Waste Policy Act of 1982, secs. 135, 141 (42
U.S.C. 10155, 10161); 44 U.S.C. 3504 note.

179. In § 150.15, revise paragraphs
(a)(7)(iii) and (a)(8) to read as follows:

■

§ 150.15

Persons not exempt.

(a) * * *
(7) * * *
(iii) Greater than Class C (GTCC)
waste, as defined in part 72 of this
chapter, in an ISFSI or an MRS licensed
under part 72 of this chapter; the GTCC
waste must originate in, or be used by,
a facility licensed under parts 50, 52, or
53 of this chapter.
(8) Greater than Class C waste, as
defined in part 72 of this chapter, that
originates in, or is used by, a facility
licensed under parts 50, 52, or 53 of this
chapter and is licensed under part 30
and/or part 70 of this chapter.
*
*
*
*
*
PART 170—FEES FOR FACILITIES,
MATERIALS, IMPORT AND EXPORT
LICENSES, AND OTHER
REGULATORY SERVICES UNDER THE
ATOMIC ENERGY ACT OF 1954, AS
AMENDED
180. The authority citation for 10 CFR
part 170 continues to read as follows:

■

Authority: Atomic Energy Act of 1954,
secs. 11, 161(w) (42 U.S.C. 2014, 2201(w));
Energy Reorganization Act of 1974, sec. 201
(42 U.S.C. 5841); 42 U.S.C. 2215; 31 U.S.C.
901, 902, 9701; 44 U.S.C. 3504 note.

181. In § 170.3, revise the definitions
for ‘‘Manufacturing License,’’ ‘‘Part 55
Reviews,’’ ‘‘Power reactor,’’ and
‘‘Special projects’’ to read as follows:

■

§ 170.3

Definitions.

*

*
*
*
*
Manufacturing license means a
license under subpart F of part 52 of this
chapter or subpart H of part 53 of this
chapter to manufacture a nuclear power
reactor(s) to be operated at sites not
identified in the license application.
*
*
*
*
*
Part 55 Reviews as used in this part
means those services provided by the
Commission to administer
requalification and replacement
examinations and tests for reactor

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Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules
operators licensed under 10 CFR part 55
or 53 of the Commission’s regulations
and employed by part 50 or 53
licensees. These services also include
related items such as the preparation,
review, and grading of the examinations
and tests.
*
*
*
*
*
Power reactor means a nuclear reactor
designed to produce electrical or heat
energy licensed by the Commission
under the authority of section 103 or
subsection 104b of the Act, and under
the provisions of §§ 50.21(b), 50.22, or
part 53 of this chapter.
*
*
*
*
*
Special projects means specific
services provided by the Commission
for which fees are not otherwise
specified in this chapter. This includes,
but is not limited to, contested hearings
on licensing actions directly related to
U.S. Government national security
initiatives (as determined by the NRC),
topical report reviews, early site
reviews, waste solidification activities,
activities related to the tracking and
monitoring of shipment of classified
matter, services provided to certify
licensee, vendor, or other private
industry personnel as instructors for 10
CFR part 55 or 53 reactor operators,
reviews of financial assurance
submittals that do not require a license
amendment, reviews of responses to
Confirmatory Action Letters, reviews of
uranium recovery licensees’ land-use
survey reports, and reviews of §§ 50.71
or 53.1545 of this chapter Final Safety
Analysis Reports. Special projects does
not include activities otherwise exempt
from fees under this part. It also does
not include those contested hearings for
which a fee exemption is granted in
§ 170.11(a)(2), including those related to
individual plant security modifications.
*
*
*
*
*
■ 182. In § 170.12, revise paragraph
(d)(1)(v) to read as follows:
§ 170.12

Payment of fees.

*
*
*
*
(d) * * *
(1) * * *
(v) 10 CFR 50.71 or 53.1545 final
safety analysis reports;
*
*
*
*
*
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*

§ 170.21

[Amended]

183. In § 170.21, in footnote 1 remove
the phrase ‘‘(e.g., 10 CFR 50.12, 10 CFR
73.5)’’ and add in its place the phrase
‘‘(e.g., 10 CFR 50.12, 10 CFR 53.080, 10
CFR 73.5)’’.
■ 184. Revise § 170.41to read as follows:
■

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§ 170.41 Failure by an applicant or
licensee to pay prescribed fees.

If the Commission determines that an
applicant or a licensee has failed to pay
a prescribed fee required in this part,
the Commission will not process any
application and may suspend or revoke
any license or approval issued to the
applicant or licensee. The Commission
may issue an order with respect to
licensed activities that the Commission
determines to be appropriate or
necessary to carry out the provisions of
this part, parts 30, 31, 32 through 35, 40,
50, 53, 61, 70, 71, 72, 73, and 76 of this
chapter, and of the Act.
PART 171—ANNUAL FEES FOR
REACTOR LICENSES AND FUEL
CYCLE LICENSES AND MATERIALS
LICENSES, INCLUDING HOLDERS OF
CERTIFICATES OF COMPLIANCE,
REGISTRATIONS, AND QUALITY
ASSURANCE PROGRAM APPROVALS
AND GOVERNMENT AGENCIES
LICENSED BY THE NRC
185. The authority citation for 10 CFR
part 171 continues to read as follows:

■

Authority: Atomic Energy Act of 1954,
secs. 11, 161(w), 223, 234 (42 U.S.C. 2014,
2201(w), 2273, 2282); Energy Reorganization
Act of 1974, sec. 201 (42 U.S.C. 5841); 42
U.S.C. 2215; 44 U.S.C. 3504 note.
■

186. Revise § 171.3 to read as follows:

§ 171.3

Scope.

The regulations in this part apply to
any person holding an operating license
for a test reactor or research reactor
issued under part 50 of this chapter, and
to any person holding an operating
license for a power reactor licensed
under 10 CFR part 50 or 53, or a
combined license issued under 10 CFR
part 52 or 53, that has provided
notification to the U.S. Nuclear
Regulatory Commission (NRC) that the
licensee has successfully completed
power ascension testing. The
regulations in this part also apply to any
person holding a materials license as
defined in this part, a Certificate of
Compliance, a sealed source or device
registration, a quality assurance program
approval, and to a Government agency
as defined in this part. Notwithstanding
the other provisions in this section, the
regulations in this part do not apply to
uranium recovery and fuel facility
licensees until after the Commission
verifies through inspection that the
facility has been constructed in
accordance with the requirements of the
license.
■ 187. In § 171.5, revise the definitions
for ‘‘Operating license,’’ and ‘‘Power
reactor’’ to read as follows:

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§ 171.5

87127

Definitions.

*

*
*
*
*
Operating license means having a
license issued under §§ 50.57 or 53.1387
of this chapter. It does not include
licenses that only authorize possession
of special nuclear material after the
Commission has received a request from
the licensee to amend its licensee to
permanently withdraw its authority to
operate or the Commission has
permanently revoked such authority.
*
*
*
*
*
Power reactor means a nuclear reactor
designed to produce electrical or heat
energy and licensed by the Commission
under the authority of section 103 or
subsection 104b of the Atomic Energy
Act of 1954, as amended, and under the
provisions of §§ 50.21(b) or 50.22, or
part 53 of this chapter.
*
*
*
*
*
■ 188. In § 171.15, revise paragraphs (a),
(b)(2)(iii), (c)(1), and (d)(1) to read as
follows:
§ 171.15 Annual fees: Non-power
production or utilization licenses, reactor
licenses, and independent spent fuel
storage licenses.

(a) Each person holding an operating
license for one or more non-power
production or utilization facilities under
10 CFR part 50 that has provided
notification to the NRC of the successful
completion of startup testing; each
person holding an operating license for
a power reactor licensed under 10 CFR
part 50 or a combined license under 10
CFR part 52, or an operating license or
combined license for a commercial
nuclear plant under 10 CFR part 53, that
has provided notification to the NRC of
the successful completion of power
ascension testing; each person holding a
10 CFR part 50 or 52, power reactor
license, or a 10 CFR part 53 commercial
nuclear plant license that is in
decommissioning or possession only
status, except those that have no spent
fuel onsite; and each person holding a
10 CFR part 72 license who does not
hold a 10 CFR part 50, 52, or 53 license
and provides notification under
§ 72.80(g) of this chapter, shall pay the
annual fee for each license held during
the Federal fiscal year in which the fee
is due. This paragraph (a) does not
apply to test or research reactors
exempted under § 171.11(b).
(b) * * *
(2) * * *
(iii) Generic activities required largely
for NRC to regulate power reactors (e.g.,
updating part 50, part 52, or part 53 of
this chapter, operating the Incident
Response Center, new reactor regulatory
infrastructure). The base annual fee for
operating power reactors does not

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Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules

include generic activities specifically
related to reactor decommissioning.
(c)(1) The FY 2022 annual fee for each
power reactor holding a 10 CFR part 50
operating license or combined license
issued under 10 CFR part 52 or part 53
that is in a decommissioning or
possession-only status and has spent
fuel onsite, and for each independent
spent fuel storage 10 CFR part 72
licensee who does not hold a 10 CFR
part 50 or part 53 operating license, or
a 10 CFR part 52 or part 53 combined
license, is $227,000.
*
*
*
*
*
(d)(1) Each person holding an
operating license for an SMR issued
under 10 CFR part 50 or part 53, or a
combined license issued under 10 CFR
part 52 or part 53, that has provided
notification to the NRC of the successful
completion startup testing, shall pay the
annual fee for all licenses held for an
SMR site. The annual fee will be
determined using the cumulative
licensed thermal power rating of all
SMR units and the bundled unit
concept, during the fiscal year in which
the fee is due. For a given site, the use
of the bundled unit concept is
independent of the number of SMR
plants, the number of SMR licenses
issued, or the sequencing of the SMR
licenses that have been issued.
*
*
*
*
*
■ 189. In § 171.17, revise paragraphs (a)
introductory text, (a)(1)(ii), and (a)(2) to
read as follows:
§ 171.17

Proration.

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who do not hold 10 CFR part 50, 52, or

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53 licenses, and materials licenses with
annual fees of $100,000 or greater for a
single fee category. The NRC will base
the proration of annual fees for
terminated and downgraded licenses on
the fee rule in effect at the time the
action is official. The NRC will base the
determinations on the proration
requirements under paragraphs (a)(2)
and (3) of this section.
(1) * * *
(ii) The annual fees for new licenses
for non-power production or utilization
facilities, 10 CFR part 72 licensees who
do not hold 10 CFR part 50, 52, or 53
licenses, and materials licenses with
annual fees of $100,000 or greater for a
single fee category for the current FY,
that are subject to fees under this part
and are granted a license to operate on
or after October 1 of a FY, are prorated
on the basis of the number of days
remaining in the FY. Thereafter, the full
annual fee is due and payable each
subsequent FY.
(2) Terminations. The base operating
power reactor annual fee for operating
reactor licensees or the annual fee for
small modular reactor licensees, who
have requested amendment to withdraw
operating authority permanently during
the FY will be prorated based on the
number of days during the FY the
license was in effect before docketing of
the certifications for permanent
cessation of operations and permanent
removal of fuel from the reactor vessel
or when a final legally effective order to
permanently cease operations has come
into effect. The spent fuel storage/
reactor decommissioning annual fee for
reactor licensees who permanently

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cease operations and have permanently
removed fuel from the site during the
FY will be prorated on the basis of the
number of days remaining in the FY
after docketing of both the certifications
of permanent cessation of operations
and permanent removal of fuel from the
site. The spent fuel storage/reactor
decommissioning annual fee will be
prorated for those 10 CFR part 72
licensees who do not hold a 10 CFR part
50, 52, or 53 license who request
termination of the 10 CFR part 72
license and permanently cease activities
authorized by the license during the FY
based on the number of days the license
was in effect before receipt of the
termination request. The annual fee for
materials licenses with annual fees of
$100,000 or greater for a single fee
category for the current FY will be
prorated based on the number of days
remaining in the FY when a termination
request or a request for a possessiononly license is received by the NRC,
provided the licensee permanently
ceased licensed activities during the
specified period. The annual fee for
non-power production or utilization
facilities will be prorated based on the
number of days remaining in the FY
when the authorization to operate the
facility has been permanently removed
from the license during the FY.
*
*
*
*
*
Dated: October 7, 2024.
For the Nuclear Regulatory Commission.
Carrie Safford,
Secretary of the Commission.
[FR Doc. 2024–23434 Filed 10–23–24; 8:45 am]
BILLING CODE 7590–01–P

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