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Rule2026-06048

Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors

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Published

March 30, 2026

Effective

April 29, 2026

Issuing agencies

Nuclear Regulatory Commission

Abstract

The U.S. Nuclear Regulatory Commission (NRC) is amending its 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 current application and licensing requirements were primarily developed to address license requests concerning light water- cooled reactors and operational requirements for those types of reactors. This final rule responds to NEIMA by creating an alternative, technology-inclusive regulatory framework to accommodate licensing of future commercial nuclear plants, including advanced reactor designs that may not employ light-water technology.

Full Text

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<title>Federal Register, Volume 91 Issue 60 (Monday, March 30, 2026)</title>
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[Federal Register Volume 91, Number 60 (Monday, March 30, 2026)]
[Rules and Regulations]
[Pages 15696-15881]
From the Federal Register Online via the Government Publishing Office [<a href="http://www.gpo.gov">www.gpo.gov</a>]
[FR Doc No: 2026-06048]

[[Page 15695]]

Vol. 91

Monday,

No. 60

March 30, 2026

Part II

Nuclear Regulatory Commission

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10 CFR Parts 1, 2, et al.

Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced
Reactors; Final Rule

Federal Register / Vol. 91 , No. 60 / Monday, March 30, 2026 / Rules
and Regulations

[[Page 15696]]

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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

AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.

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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is amending its
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 current application and licensing requirements were
primarily developed to address license requests concerning light water-
cooled reactors and operational requirements for those types of
reactors. This final rule responds to NEIMA by creating an alternative,
technology-inclusive regulatory framework to accommodate licensing of
future commercial nuclear plants, including advanced reactor designs
that may not employ light-water technology.

DATES: This final rule is effective on April 29, 2026.

ADDRESSES: 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:
<bullet> Federal Rulemaking Website: Go to <a href="https://www.regulations.gov">https://www.regulations.gov</a> and search for Docket ID NRC-2019-0062. Address
questions about NRC dockets to Helen Chang; telephone: 301-415-3228;
email: <a href="/cdn-cgi/l/email-protection#327a575e575c1c715a535c55725c40511c555d44">[email&#160;protected]</a>. For technical questions, contact the
individuals listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
<bullet> NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly available documents online in the
ADAMS Public Documents collection at <a href="https://www.nrc.gov/reading-rm/adams.html">https://www.nrc.gov/reading-rm/adams.html</a>. To begin the search, select ``Begin 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 <a href="/cdn-cgi/l/email-protection#f2a2b6a0dca097819d87809197b29c8091dc959d84">[email&#160;protected]</a>. For the convenience of the reader,
instructions about obtaining materials referenced in this document are
provided in the ``Availability of Documents'' section.
<bullet> 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
<a href="/cdn-cgi/l/email-protection#ecbca8bec2be899f83999e8f89ac829e8fc28b839a">[email&#160;protected]</a> 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.

FOR FURTHER INFORMATION CONTACT: Nicole Fields, Office of Nuclear
Material Safety and Safeguards, telephone: 630-829-9570, email:
<a href="/cdn-cgi/l/email-protection#f8b6919b97949dd6be919d949c8bb8968a9bd69f978e">[email&#160;protected]</a> and Anders Gilbertson, Office of Nuclear Reactor
Regulation, telephone: 301-415-1541, email: <a href="/cdn-cgi/l/email-protection#6c2d0208091e1f422b05000e091e181f03022c021e0f420b031a">[email&#160;protected]</a>.
Both are staff of the U.S. Nuclear Regulatory Commission, Washington,
DC 20555-0001.

SUPPLEMENTARY INFORMATION: This rulemaking is separate from the NRC's
comprehensive review and reform of its regulations, including those
governing transportation, in accordance with Executive Order (E.O.)
14300, ``Ordering the Reform of the Nuclear Regulatory Commission'' (90
FR 22587; May 29, 2025). The rulemakings associated with that effort
will comprehensively reexamine NRC requirements. While there could be
additional revisions to part 53, ``Risk-Informed, Technology-Inclusive
Regulatory Framework for Commercial Nuclear Plants,'' of title 10 of
the Code of Federal Regulations (10 CFR) as a result of these future
rulemakings, the NRC is moving forward with publication of this final
rule at this time because it is a deregulatory action of high interest
for stakeholders that was in progress before the issuance of E.O.
14300.

Executive Summary

A. Need for the Regulatory Action

On January 14, 2019, the President signed the NEIMA into law (Pub.
L. 115-439). NEIMA section 103(a)(4) directs the NRC to ``complete a
rulemaking to establish a technology-inclusive, 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 of the Code of Federal Regulations (10 CFR) (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 10 CFR part
53 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 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'' is
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 light water-cooled reactors and operational
requirements for those types of reactors. This final rule responds to
NEIMA by creating an alternative, technology-inclusive regulatory
framework to accommodate licensing of future commercial nuclear plants,
including advanced reactor designs that may not employ light-water
technology. The new alternative requirements and implementing guidance
adopt technology-inclusive approaches and use risk-informed and
performance-based techniques to ensure an equivalent level of safety to
that of operating commercial nuclear plants while providing optionality
and flexibility for licensing and regulating a variety of technologies
and designs for commercial nuclear reactors.

[[Page 15697]]

B. Major Provisions

Major provisions of this final rule, supported by accompanying
guidance, include the following:
<bullet> A new alternative technology-inclusive, risk-informed,
performance-based framework that includes requirements for licensing
and regulating nuclear plants during the various stages of their life
cycles.
<bullet> A new alternative technology-inclusive, 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.
<bullet> A new alternative technology-inclusive 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 final regulatory analysis to determine the
expected quantitative costs and benefits of this final rule and
associated guidance as well as qualitative factors to be considered in
the NRC's rulemaking decision. The conclusion from the analysis is that
this final rule and associated guidance would result in net averted
costs to the industry and the NRC of $152 million using a 7-percent
discount rate and $203 million using a 3-percent discount rate. The
annualized averted costs at a 7-percent discount rate are approximately
$1.64 million per year to the NRC and $9.1 million per year to
industry, or net annualized averted costs of approximately $10.7
million, over the 66-year analysis period. The number of future
applicants was chosen conservatively, based on information known to the
NRC; with each additional applicant beyond those included in the
regulatory analysis, this final rule becomes even more cost-beneficial.
The final regulatory analysis also considers qualitative factors
such as greater regulatory stability, predictability, and clarity to
the licensing process. These benefits would result, for example, from
incorporating advances in probabilistic risk assessment (PRA) and other
risk-informed analyses into the regulatory framework. 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 final regulatory analysis
(available in the NRC's Agencywide Documents Access and Management
System (ADAMS) Accession No. ML26042A230).

Table of Contents

I. Background
NRC Advanced Reactor Readiness
II. Discussion
A. Objective and Applicability
B. Need for Changes to the Existing Regulatory Framework
C. 10 CFR Part 53 Framework
III. Part 53 Framework
Subpart A--General Provisions
A. Discussion of Definitions in 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
IV. Changes to Other Parts of 10 CFR Chapter I
10 CFR Part 26
A. Introduction
B. Changes to Part 26, Subparts A Through E and I
C. Requirements for Part 26, Subpart M
D. Changes to Part 26, Subpart N
E. 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
C. Appendix E to Part 50: Emergency Planning and Preparedness
for Production and Utilization Facilities
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
V. Opportunities for Public Participation
VI. Public Comment Analysis
VII. Regulatory Flexibility Certification
VIII. Regulatory Analysis
IX. Backfitting and Issue Finality
X. Cumulative Effects of Regulation
XI. Plain Writing
XII. Environmental Assessment and Final Finding of No Significant
Environmental Impact
XIII. Paperwork Reduction Act
XIV. Executive Orders
A. Executive Order 12866: Regulatory Planning and Review (as
Amended by Executive Order 14215: Ensuring Accountability for All
Agencies)
B. Executive Order 14154: Unleashing American Energy
C. Executive Order 14192: Unleashing Prosperity Through
Deregulation
D. Executive Order 14270: Zero-Based Regulatory Budgeting To
Unleash American Energy
XV. Congressional Review Act
XVI. Criminal Penalties
XVII. Voluntary Consensus Standards
XVIII. Availability of Guidance
XIX. Availability of Documents

I. Background

The NRC is amending its regulations by adding an alternative risk-
informed, performance-based, and technology-inclusive regulatory
framework as an option for the licensing and regulation of future
commercial nuclear plants. This section discusses previous activities
that have led to the development of this final rule.

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 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

[[Page 15698]]

improve risk-informed and performance-based approaches within the
agency (e.g., the Commission's PRA policy statement, ``Use of
Probabilistic Risk Assessment Methods in Nuclear Regulatory
Activities,'' dated August 16, 1995 (60 FR 42622)).
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. These activities leveraged
previous work described in NUREG-1860, ``Feasibility Study for a Risk-
Informed and Performance-Based Regulatory Structure for Future Plant
Licensing,'' published in 2007. One key element of these efforts was
the Licensing Modernization Project (LMP), a cost-shared initiative led
by nuclear utilities and supported by DOE. The LMP methodology is a
technology-inclusive, risk-informed, and performance-based methodology
developed for non-LWR designs. The LMP methodology 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
methodology 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 methodology
activities led to the publication and submittal of Nuclear Energy
Institute (NEI) 18-04, Revision 1, ``Risk-Informed Performance-Based
Technology-Inclusive Guidance for Non-Light 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 methodology.
The NRC endorsed the principles and methodology in NEI 18-04, with
clarifications, in 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 Non-Light-Water Reactors.'' The NRC
staff sought Commission approval of the use of the LMP methodology 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 methodology 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 non-
LWRs in Staff Requirements Memorandum (SRM) SRM-SECY-19-0117, dated May
26, 2020. Although the LMP methodology is technology-inclusive, the
industry and NRC staff initially focused the LMP methodology's
applicability on non-LWRs, both for efficiency and to support 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, dated April
13, 2020, the NRC staff developed part 53 by building upon recent and
ongoing activities such as the LMP methodology described in SECY-19-
0117. Such an approach supports implementing the NEIMA direction to
establish a technology-inclusive framework as well as the 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. The LMP methodology highlights the role of
PRA in risk-informed and performance-based approaches to identifying
enhanced safety margins that can be used to justify operational
flexibilities. The part 53 framework is largely based on the
methodology described in SECY-19-0117 and includes a prominent role for
PRA, other systematic risk evaluations (SREs), or a combination
thereof.

II. Discussion

A. Objective and Applicability

The NRC is adding a new, alternative part to its regulations that
sets out a risk-informed, technology-inclusive framework for the
licensing and regulation of commercial nuclear plants. This new
approach achieves 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 final rule
adds 10 CFR part 53; subpart M, ``Fitness-for-Duty Programs for
Facilities Licensed Under 10 CFR part 53,'' to part 26; Sec. 73.100,
``Technology-inclusive requirements for physical protection of licensed
activities at commercial nuclear plants against radiological
sabotage,'' Sec. 73.110, ``Technology-inclusive requirements for
protection of digital computer and communication systems and
networks,'' and Sec. 73.120, ``Access authorization program for
commercial nuclear plants,'' as well as makes 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 presents

[[Page 15699]]

challenges because the existing regulatory framework has evolved
primarily to address the LWR designs that compose the current operating
fleet. The NRC has had many interactions with designers of various
reactor technologies under development, sometimes collectively referred
to as advanced 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 early efforts to
establish a technology-inclusive 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 technology-inclusive 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 has evaluated
the ADVANCE Act, including how NRC regulations, such as 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 microreactors, nuclear reactor license
application reviews, and nuclear fuel. Finally, in 2025, the President
signed E.O. 14300, ``Ordering the Reform of the Nuclear Regulatory
Commission,'' which builds on the provisions in the ADVANCE Act. E.O.
14300 will complement this rulemaking by providing additional
mechanisms for streamlining the agency's efforts to provide an
efficient licensing pathway for advanced reactors.
The requirements in part 53 support a wide variety of potential
commercial nuclear reactor technologies. 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 Statements. The
Commission reiterated its safety 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 light-water reactors. New reactors with these
enhanced margins and safety features should have greater operational
flexibility than current reactors.

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. As a
result, the existing regulations have benefited from a focused and
tailored treatment of safety issues as issues arose and evolved. In
contrast, part 53 is being developed to accommodate technologies that,
in some cases, lack significant operating experience. This lack of
operating experience makes it challenging to develop technology-
inclusive regulatory requirements when it is less well-known which
issues may be more or less important to safety for any given set of
technologies. To address these challenges, the NRC drew on well-
developed approaches to licensing to produce a technology-neutral and
robust regulatory framework. The regulatory framework uses PRAs, other
SREs, or a combination thereof, to assess risks and focus on the issues
most important to safety, help establish technical requirements, and
manage operations. The framework builds on the LMP methodology, which
is a technology-inclusive approach to licensing that leverages risk
insights to provide applicants with significant design and operation
flexibilities.

C. 10 CFR Part 53 Framework

This final 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. The major features of this final rule include the following:
(1) Technology-inclusiveness. This rule provides a broad and
flexible regulatory framework that can be used for any reactor
technology, any size reactor, and any reactor end use.
(2) Risk-informed framework to support safety-focused decision-
making. Part 53 provides a holistic, risk-informed framework that
offers substantial flexibility in leveraging safety margins and
focusing on design features and programmatic controls important to
protecting public health and safety. The framework allows for explicit
consideration of risk through the use of PRAs or other SRE techniques,
or a combination thereof, to generate risk insights, and to assess and
manage those risks. This approach departs from traditional
deterministic methods, notably the use of the single-failure criterion,
by enabling applicants to propose comprehensive risk metrics and
associated risk performance objectives, appropriate systematic risk
assessment techniques, and to demonstrate how their design and
associated programmatic controls protect public health and safety.
(3) Performance-based approach. Part 53 is a performance-based
framework that provides flexibility in establishing appropriate high-
level safety objectives and demonstrating how a reactor design or
specific commercial nuclear plant meets those objectives. Rather than
prescribing specific methods or processes, the performance-based
approach in part 53 promotes efficiency and innovation by allowing
applicants to propose design features to meet safety objectives and
achieve safety outcomes. This will support novel concepts such as
leveraging functional containment concepts, alternative siting criteria
for commercial nuclear reactors in relation to population centers,
reduced staffing

[[Page 15700]]

levels, and remote operations, while eliminating traditional,
prescriptive requirements, such as general design criteria and aircraft
impact assessments.
(4) Licensing pathways that accommodate a broad spectrum of design
maturities and deployment models. Part 53 provides several licensing
options for applicants to choose from to meet their deployment model or
business case needs, including the licenses, certifications, and
approvals provided by parts 50 and 52. This final rule provides
additional flexibility for manufacturing licenses (MLs), including the
possible factory loading of fuel into manufactured reactors with
appropriate features to prevent criticality for deployment to another
location for operation.
(5) Operator licensing. Part 53 introduces the concept of self-
reliant-mitigation facilities and the use of generally licensed reactor
operators (GLROs) for those facilities. The allowance for GLROs
provides flexibility for the types and locations of staffing needed
under part 53.
(6) Efficiency. Part 53 provides opportunities to improve
regulatory efficiency by including provisions for licensing first-of-a-
kind proposals as well as provisions that benefit those proposing
standardized and repetitious applications. Part 53 provides finality to
designs for which an operating license has been issued to improve its
incorporation into a standardized design approval or certification.
Part 53 also provides for a risk-informed approach for managing plant
equipment and programmatic controls that reduce the future need for
regulatory approvals.
(7) Codes and standards. Part 53 does not incorporate by reference
specific codes and standards as is done in Sec. 50.55a, ``Codes and
standards.'' Instead, part 53 allows the use of generally accepted
codes and standards to be tailored to the assessed safety significance
of SSCs, such as the use of non-nuclear codes and standards for SSCs
composed of commercial grade components.
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 systems-engineering style approach to the
design, licensing, operation, and ultimately decommissioning of future
commercial nuclear plants. Organizing requirements in this manner also
supports performance-based 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 complement the design and analysis
requirements in subpart C. The performance-based approach adopted in
part 53 also includes regulatory requirements that 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.
Subpart M of part 26 is new and is 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. Subpart M
of part 26 is 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
requirements are not entirely equivalent because current subpart K of
part 26 only applies during construction of the commercial nuclear
plant, whereas subpart M of part 26 applies during construction,
operation, and decommissioning. Furthermore, subpart M of part 26
allows 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.
Revisions to part 73 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 final rule makes conforming changes throughout 10
CFR chapter I, by adding ``and part 53'' where appropriate to account
for the addition of part 53.

III. Part 53 Framework

Subpart A--General Provisions

Subpart A provides the general provisions applicable to all
applicants and licensees that are 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
AEA) and title II of the Energy Reorganization Act of 1974 (88 Stat.
1242). Subpart A includes 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 are largely equivalent to the general
requirements in part 50 that are applicable to all part 50 applicants
and licensees (specifically, Sec. Sec. 50.1 through 50.13) but
reference the corresponding regulations in part 53 in place of
references to part 50.
A. Discussion of Definitions in Part 53
This final rule includes a definition section in Sec. 53.020. The
definitions of most terms in Sec. 53.020 are equivalent to the
corresponding terms defined in: (1) Sec. Sec. 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 Sec. 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
nuclear fission reactor or fusion machine, including a prototype plant
(as defined in sections 50.2 and 52.1 of 10 CFR (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,

[[Page 15701]]

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'' is 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 <a href="https://www.regulations.gov">https://www.regulations.gov</a>.
The NRC allows the use of part 53 by any ``commercial nuclear
plant.'' The use of the term ``plant'' versus ``reactor,'' as used in
existing regulations (i.e., Sec. 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 Sec. 50.2. However, the phrase ``in a self-
supporting chain reaction'' is not included in the definition of
Commercial nuclear plant to enable applying part 53 to accelerator
driven systems that use special nuclear material (SNM) but that do not
involve self-sustaining chain reactions. Relatedly, ``Utilization
facility'' is also defined in Sec. 53.020 based on the definition of
that term in Sec. 50.2 and refers to a ``Commercial nuclear plant'' as
defined in Sec. 53.020.
The definition of ``Construction'' is different from the definition
in Sec. 50.10. Because the regulatory framework in part 53 uses risk-
informed, less prescriptive, and performance-based requirements as
compared to part 50, the part 53 definition takes a different approach
in determining what activities are prohibited without an NRC license.
Under the part 53 approach, the definition of ``Construction''
specifies a variety of activities that are applicable to safety-related
(SR) and non-safety-related but safety-significant (NSRSS) SSCs and are
credited or relied upon for demonstrating compliance with safety
criteria defined in subpart B of part 53 as well as SSCs necessary to
comply with part 73 and onsite emergency facilities necessary to comply
with Sec. 53.855. By listing the activities for SR and NSRSS SSCs that
are credited or relied upon for demonstrating compliance with safety
criteria defined in subpart B, this definition describes activities
related to SSCs subject to some sort of special treatment, as that term
is defined in Sec. 53.020. These special treatment requirements, which
include quality assurance, design criteria, and programmatic controls,
apply to safety-related SSCs and the set of non-safety-related SSCs for
which a license is required to authorize construction activities. The
latter category includes a facility's NSRSS SSCs. The non-safety-
significant SSCs not subject to special treatment and NSRSS SSCs for
which special treatments are limited to operational controls are, in
general, identified as ``commercial grade'' and may be designed,
procured, and installed in accordance with the usual practices employed
for industrial plants. Importantly, under the part 53 definition, an
SSC that falls outside the definition of construction may still be
subject to the NRC's statutory authority during operations. In view of
the foregoing, the definition of ``Construction'' in Sec. 53.020 is
consistent with the provisions of the AEA related to construction
permits, while simultaneously allowing activities related to SSCs that
are commercial grade but which could still be subject to the NRC's
jurisdiction during operations. This definition also includes the
listed activities which are for SSCs necessary to comply with part 73
or onsite emergency facilities necessary to comply with Sec. 53.855.
The inclusion of the listed activities which are for these SSCs is
consistent with Sec. 50.10(a)(1)(v) and (vii), which include
activities for corresponding SSCs. Including these activities in the
definition of ``Construction'' is appropriate because, in both
instances, part 53 points back to the relevant existing frameworks in
part 73 and the relevant part 50 requirements, respectively, rather
than creating an entirely new framework. Section 53.020 also adds
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 non-safety-significant SSCs);
performance metrics (e.g., safety criteria and functional design
criteria); and special treatment.
The regulation defines ``Safety criteria'' in terms of the plant-
level performance-based metrics that are provided in Sec. Sec. 53.210
and 53.220. The term ``Functional design criteria'' is 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'' refers to those SSCs needed to
meet the safety criteria in Sec. 53.210. The term ``Non-safety-related
but safety-significant SSCs'' means those SSCs that are not SR because
they are not relied upon to perform any function necessary to
demonstrate compliance with Sec. 53.210 but warrant special treatment
because they are relied on to achieve adequate defense in depth or
perform risk-significant functions. The term ``Non-safety-significant
SSCs'' means those SSCs that are not SR or NSRSS.
The term ``Programmatic controls'' means administrative measures
that govern human action in implementing programs and operating,
monitoring, and maintaining SSCs and equipment of a commercial nuclear
plant.
The terms ``Design-basis accidents,'' ``Anticipated event
sequences,'' ``Unlikely event sequences,'' and ``Very unlikely event
sequences'' are defined to be different types of ``Licensing-basis
events'' and are also largely equivalent to the LMP methodology's
definitions of DBAs, anticipated operational occurrences (AOOs),
design-basis events (DBEs), and beyond-design-basis events,
respectively. The term ``Design-basis accidents'' is 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 are derived from the
unlikely event sequences from the PRA, a type of SRE, other SREs, or a
combination thereof, and then analyzed in a conservative approach by

[[Page 15702]]

prescriptively assuming that only SR SSCs are available to mitigate
postulated accident scenarios. Within the LMP methodology, event
sequences with mean frequencies of 1x10\-2\/plant-year and greater are
classified as anticipated event sequences. Within the LMP methodology,
infrequent event sequences with mean frequencies of 1x10\-4\/plant-year
to 1x10\-2\/plant-year are classified as unlikely event sequences.
``Very unlikely event sequences'' are less likely to occur than
unlikely event sequences. Within the LMP methodology, rare event
sequences with frequencies of 5x10\-7\/plant-year to 1x10\-4\/plant-
year are classified as very unlikely event sequences. While the
terminology for these event sequences creates some differences between
part 53 and the LMP methodology, part 53 uses 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-design-basis accidents terminology, these definitions
seek to clarify the event categories in part 53. Finally, although the
term ``event sequence'' is often used in the context of a PRA, that
term is used generically in part 53 and does not imply the use of a
specific type of SRE, such as a PRA. The sections of this preamble
related to subparts B and C provide additional discussion of LBEs.
Section 53.020 includes a definition of ``Special treatment'' to
explain that it means those requirements, such as quality assurance
(QA), design criteria, and programmatic controls, that are taken beyond
the procurement, installation, and maintenance of commercial grade
products. Routine commercial practices may include the use of selected
consensus codes and standards that are cited in applications to support
the identification of special treatments that may go beyond what would
otherwise be required by those selected commercial codes and standards.
The special treatments increase confidence 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 subpart C. Structures, systems, and components
designated as SR 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 Sec. 53.210.
To maintain alignment with definitions in part 52, the NRC has
added a definition of early site permit (ESP). The NRC proposed
definitions for ``Consensus code or standard'' and ``probabilistic risk
assessment'' but is not including a definition for these terms in this
final rule because these terms were determined not to be essential for
the framework and including the definitions could introduce issues with
consistency given alternative definitions developed by other
organizations.
B. Other General Provisions
Section 53.040 governs written communications and how applications
and other required information must be submitted to the NRC. These
requirements are equivalent to those in Sec. 50.4.
Section 53.050 establishes 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 are
equivalent to those in Sec. 50.5.
Section 53.060 prohibits 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 for an NRC license, permit, DC, or design approval for
engaging in certain protected activities. Section 53.060 also
prescribes a procedure for seeking a remedy for employees who believe
they have been discriminated against for engaging in such protected
activities. These requirements are equivalent to those in Sec. Sec.
50.7 and 52.5.
Section 53.070 governs the completeness and accuracy of information
provided to the NRC. These requirements are equivalent to those in
Sec. Sec. 50.9 and 52.6.
Section 53.080 governs exemptions from the requirements of the
regulations in part 53. These requirements are equivalent to those in
Sec. Sec. 50.12 and 52.7.
Paragraphs (a) through (d) of Sec. 53.090 establish requirements
for standards that the NRC will consider in determining whether a
construction permit (CP), operating license (OL), ESP, combined
license, or ML under part 53 will be issued to an applicant. These
requirements are equivalent to those in Sec. Sec. 50.40, 50.42, 50.43
and 50.22, respectively. Requirements equivalent to those in Sec. Sec.
50.41 and 50.21 are not included in part 53 because they apply to Class
104 licenses, and part 53 does not apply to those licenses.
Section 53.100 requires that no license issued under part 53 may
cover activities that are not under or within the jurisdiction of the
United States. These requirements are equivalent to those in Sec.
50.53.
Section 53.110 states that licensees and applicants are not
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 are equivalent to those in Sec. 50.13.
Section 53.115 establishes requirements for rights related to SNM.
These requirements are equivalent to those in Sec. 50.54(b) and (c).
Section 53.117 establishes 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 are equivalent to those in Sec. 50.54(d).
Section 53.120 establishes requirements for information collection
requirements that have received Office of Management and Budget (OMB)
approval. These requirements are equivalent to those in Sec. 50.8.

Subpart B--Technology-Inclusive Safety Requirements

Subpart B, ``Technology-Inclusive Safety Requirements,'' provides
technology-inclusive safety criteria that serve as performance
standards for the subsequent performance-based requirements used
throughout part 53. Subsequent subparts 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 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
is built upon the safety criteria in Sec. 53.210, ``Safety criteria
for design-basis accidents.'' The more detailed requirements for those
SSCs are then 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 are governed
by subpart E, ``Construction and Manufacturing Requirements,'' and
subpart F, ``Requirements for Operation.''
Requirements for NSRSS SSCs warranting special treatment are

[[Page 15703]]

likewise determined under Sec. 53.220, ``Safety criteria for
licensing-basis events other than design-basis accidents,'' in subpart
B and Sec. 53.460, ``Safety categorization and special treatment,'' in
subpart C. Regulatory requirements related to the NSRSS SSCs are
distinguished from the regulatory requirements for SR SSCs throughout
part 53. Part 53 affords more flexibility to applicants and licensees
regarding how NSRSS SSCs are used in the design and maintained during
plant operations, as compared to SR SSCs.
The collective set of performance-based requirements in part 53 are
sufficient, if met, for the NRC to make the findings required to grant
an application for a utilization facility under section 182 of the AEA
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 is 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
minimum. The requirements throughout part 53 that support demonstrating
compliance with Sec. 53.220 are 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 AEA.
Consistent with historical practice, sections 182 and 161 of the
AEA are cited as authorizing legislation within this final rule.
However, specific language from the AEA is not incorporated into the
safety objectives or safety criteria in part 53. This is because, again
consistent with historical practice, the NRC is not defining ``adequate
protection'' through the individual safety requirements in part 53.
Rather, part 53 enables the NRC to make its required findings under the
AEA 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 provides safety criteria for DBAs that are required
to be identified under Sec. 53.240 and analyzed under Sec. 53.450(f)
in subpart C of part 53. Subsequent sections in part 53 require that
the SSCs relied upon to demonstrate compliance with the criteria in
Sec. 53.210 be classified as SR. The use of SR SSCs and the 25 rem
reference values for potential radiological consequences aligns with
traditional deterministic approaches for LWRs from Sec. Sec. 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 Sec. 50.34(a)(1)(ii)(D)(1) and Sec.
52.79(a)(1)(vi)(A) is included in Sec. 53.210 to explain that the use
of the 25 rem value 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 that is 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 provides a logical structure supporting
the identification and treatment of SR SSCs and establishing the
corresponding functional design criteria for those SSCs.
Section 53.220 provides safety criteria for LBEs other than DBAs
that are required to be identified under Sec. 53.240 and analyzed
under Sec. 53.450(e) in subpart C. Whereas Sec. 53.210 and the
related requirements for SR SSCs provide that a defined success path
exists for DBAs, the safety criteria for LBEs other than DBAs establish
the connections between SSC design, human actions, and programmatic
controls and a broader set of potential internal and external hazards.
These safety criteria also address defense-in-depth matters such as a
balanced consideration of prevention and mitigation.
The safety criterion in Sec. 53.220(b) includes 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 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
pre-established, indicative values of the comprehensive risk metrics
that are used as part of risk-informed decision-making. The methodology
for developing and using proposed comprehensive risk metrics and
associated risk performance objectives is defined by the requirements
for analyses in Sec. 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 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 do 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, is one of
several performance standards used in subpart B. The 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,

[[Page 15704]]

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
Non-Light-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 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 Sec. 53.220(b). The requirement for comprehensive
risk metrics, in combination with the other requirements in subparts B
and C, brings 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 provides a logical performance
objective to support the risk management approaches in the various
subparts comprising part 53.
The Commission stated in the introduction of the Safety Goals
Policy Statement that improvements to then-current 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 Sec. 53.220(b) can 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 final rule
establish acceptable risks, not acceptable deaths.
Applicants under 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 those previously
discussed, including the use of surrogate measures for use in specific
analyses to satisfy the requirements in Sec. 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 will, as appropriate, review novel approaches for
comprehensive risk metrics and associated risk performance objectives
proposed by applicants, industry organizations, or standard development
organizations and will engage stakeholders during the development of
the related regulatory guidance or specific licensing actions.
Section 53.230 requires safety functions needed to ensure that the
safety criteria under Sec. Sec. 53.210 and 53.220 can be met if an
assumed LBE were to occur at a commercial nuclear plant. Section 53.230
specifies 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) can 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 final 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.
Section 53.240 requires 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 Sec. 53.450 are
below the safety criteria identified under Sec. Sec. 53.210 and 53.220
and that SSCs, personnel, and programs address the safety functions
from Sec. 53.230. Including a high-level performance requirement
related to the identification of LBEs to address appropriate risk-
informed combinations of malfunctions of plant SSCs, human errors,
facility hazards, and the effects of external hazards and analysis
thereof in subpart B reflects the historical and continuing importance
of evaluating unplanned events as part of the licensing of commercial
nuclear plants. Section 53.240 requires identification and analysis of
LBEs under Sec. 53.450 using a PRA, other SREs, or a combination
thereof. An example of acceptable methods of using PRAs to identify and
assess LBEs is the methodology in RG 1.233, as discussed in RG 1.254,
``Technology-Inclusive Identification of Licensing Events for
Commercial Nuclear Plants.''
Section 53.250 establishes 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 necessitates this
specific requirement to ensure defense in depth is provided. The
requirements in this section state that no single engineered design
feature, human action, or programmatic control, no matter how robust,
should be exclusively relied upon to address the range of LBEs other
than DBAs. The requirement under Sec. 53.250(c) is different from the
single failure criterion described in appendix A to part 50. The Sec.
53.250(c) requirement does not allow the safety analysis to exclusively
rely upon a

[[Page 15705]]

single engineered design feature, human action, or programmatic control
to address the range of LBEs other than DBAs (i.e., ranging from very
unlikely event sequences to anticipated event sequences). In contrast,
the single failure criterion under appendix A to part 50 relates, in
part, to the failure of a component to perform its intended safety
function, regardless of whether that component was exclusively relied
upon to address the range of LBEs. This means the requirement under
Sec. 53.250(c) does not strictly disallow single failures, as defined
in appendix A to part 50, because a component could experience such a
single failure and, if it is not otherwise exclusively relied upon to
address the range of LBEs other than DBAs, its failure alone does not
preclude being able to satisfy Sec. 53.250(c). In that regard, Sec.
53.250 allows for greater flexibility such that other measures could be
taken to ensure appropriate defense in depth without needing to
accommodate single failures, as defined in appendix A to part 50. The
phrase ``engineered design feature'' does not preclude the possible
crediting of inherent characteristics within the design and analysis
for commercial nuclear reactors. While defense in depth is only
assessed for LBEs other than DBAs, the need to ensure dedicated success
paths for DBAs contributes to the overall defense in depth for each
commercial nuclear plant under part 53.
Section 53.260 governs normal operations and establishes a level of
safety based on requirements in 10 CFR part 20, ``Standards for
Protection Against Radiation,'' which limit doses to members of the
public and dose rates in unrestricted areas.
Section 53.270 provides for the protection of plant workers and
establishes a level of safety based on requirements in 10 CFR part 20,
which limit occupational dose.

Subpart C--Design and Analysis Requirements

This subpart provides requirements for the design of commercial
nuclear plants and the supporting analyses, including the analyses of
LBEs, to demonstrate that the performance standards in subpart B can be
satisfied. The sections within subpart C reflect the overall hierarchy
throughout part 53, which covers: (1) plant-level safety criteria
(Sec. Sec. 53.210 and 53.220); (2) safety functions (Sec. 53.230)
needed to demonstrate compliance with the safety criteria; (3) design
features (Sec. 53.400), human actions, and programmatic controls
needed to fulfill the safety functions; and (4) functional design
criteria (Sec. Sec. 53.410 and 53.420) that must be defined for each
design feature relied upon to demonstrate the safety criteria
(Sec. Sec. 53.210 and 53.220) are met. Subpart C also contributes to
the logic and structure of part 53 by distinguishing between SR SSCs
and NSRSS SSCs and licensee-controlled 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 Sec. 53.210. NSRSS 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 establishes a requirement that design features be
provided for each commercial nuclear plant to satisfy the safety
criteria and fulfill safety functions from subpart B during LBEs. Other
sections in subpart C, 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 requires that functional design criteria be defined
for safety-related design features relied upon to demonstrate that the
consequences from DBAs would be below the criteria in Sec. 53.210
through analyses performed under Sec. 53.450(f), which includes
insights from both PRAs and deterministic analyses. Other sections
within part 53 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 consideration of human actions and programmatic controls
in the identification of special treatments associated with the design
of SR SSCs will contribute to ensuring that a commercial nuclear plant
licensed under part 53 would meet the safety criteria in Sec. 53.210.
Section 53.415 requires 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, Sec. 53.415 requires
that SR SSCs remain capable of performing the safety functions stated
in Sec. 53.230 for which they are credited up to the design-basis
external hazard levels as determined under Sec. 53.510. As used in
Sec. 53.415 and subpart D of part 53, a hazard level refers 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 will 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 requires 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 are described in Sec.
53.450(e), which requires that those events be identified and assessed
using a PRA, other SREs, or a combination thereof, together with other
generally accepted approaches for systematically evaluating engineered
systems. The SSCs determined to be safety significant (i.e., either SR
or NSRSS) will have associated special treatment requirements as
specified in Sec. 53.460. Special treatment is defined in subpart A of
part 53 and generally refers to measures (e.g., quality assurance,
testing, monitoring) taken beyond normal commercial practices related
to the procurement, installation, and maintenance of commercial grade
products to provide confidence that SR and NSRSS SSCs will perform
under the service conditions and with the reliability assumed in the
analysis under Sec. 53.450(e) and will comply with the applicable
functional design criteria. Such normal commercial practices include
the use of consensus codes and standards, as identified in an
application, to support the identification of special treatments that
include measures that may go beyond the use of commercial codes and
standards. 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 safety criteria in subpart B. Therefore,
designers and licensees for commercial nuclear plants are 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 Sec. 53.220(b).

[[Page 15706]]

Section 53.425 establishes 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 provides similar requirements for design features and
related functional design criteria for protection of plant workers to
meet the safety criteria in part 20. Section 53.425 provides applicants
and licensees flexibility to define design objectives for design
features related to controlling liquid, gaseous, and solid wastes as
required under part 20. The design objective will assist designers,
applicants, and licensees in performing the evaluations of possible
reductions in public dose from routine effluents when considering costs
and other factors.
The requirements in Sec. Sec. 53.425 and 53.430 for design
features and functional design criteria to support radiation protection
activities have parallels in existing regulations such as Sec.
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 addresses various design requirements that warrant
specific mention to ensure that the design features required by Sec.
53.400 comply with the functional design criteria required by
Sec. Sec. 53.410 and 53.420. These requirements will be met through
design practices, consideration of testing and operating experience,
and various assessments of LBEs and other potential challenges to
commercial nuclear plants. Discussion of some of the key design
requirements included in section 53.440 follow.
(1) Sec. 53.440(a): An essential element to ensuring a proposed
design can comply with the performance criteria in part 53 is that the
ability of design features to fulfill their safety functions is
demonstrated by a combination of analyses, test programs, prototype
testing, and operating experience. This requirement closely aligns with
the language in Sec. 50.43(e) and is included in part 53 as the same
foundational requirement. In addition, Sec. 53.440(a) requires 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 requirement corresponds to the existing
requirement under Sec. 50.34(f)(3)(i) that was developed in response
to the 1979 accident at Three Mile Island Nuclear Generating Station.
(2) Sec. 53.440(b): The design and licensing of commercial nuclear
plants should use generally accepted consensus codes and standards for
design features classified as safety-related. 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 will 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.
(3) Sec. 53.440(c): The design requirements in subpart C require
the materials used for SR and NSRSS SSCs to be qualified for their
service conditions over the design life of the SSC as appropriate to
satisfy the special treatments established for the SSC under Sec.
53.460.
(4) Sec. 53.440(d): The requirements in Sec. 53.440 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.
(5) Sec. 53.440(e) and (f): The design requirements in subpart C
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.
Under Sec. 53.440(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.
This approach ensures considerations are given for safety and security
together throughout the plant's lifetime, including the design process
and prior to implementing changes to plant configurations, to ensure
risks are effectively managed. The implementation of a security
strategy and design features early in the design process has the
potential to be more efficient and cost-effective rather than
implementing these features after the plant has been designed and
constructed.
(6) Sec. 53.440(g) and (h): Specific design requirements will
ensure that commercial nuclear reactors under part 53 have the
capability to achieve and maintain subcriticality and long-term
cooling. The requirements are 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.
(7) Sec. 53.440(i): The design, analysis, and development of
programmatic controls under part 53 will consider the number of reactor
units and other significant inventories of radioactive materials
contributing to the risks to public health and safety. This reflects
the definition of ``Commercial nuclear plant'' in subpart A and
reinforces that the evaluation of LBEs is performed on a plant-wide
basis. This aspect of part 53 is different from parts 50 and 52, which
generally define safety requirements on the assumption of events
involving only individual reactor units.
(8) Sec. 53.440(k): The inclusion of a specific requirement for
design features and related functional design criteria, including
associated programmatic controls or a combination thereof, 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 is 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.
(9) Sec. 53.440(l): These provisions 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 Sec. 20.1406.
(10) Sec. 53.440(m): This design requirement provides a
technology-inclusive equivalent to the requirements in Sec. 50.68 by
including options for commercial nuclear plants to either have a
monitoring system capable of detecting a criticality as described in
Sec. 70.24 or to have restrictions on SNM

[[Page 15707]]

handling and storage that would prevent inadvertent criticality events.
(11) Sec. 53.440(n): The design needs to reflect state-of-the-art
human factors principles for safe and reliable performance in all
settings that human activities are expected for performing or
supporting the continued availability of plant safety or emergency
response functions.
One notable exclusion from the design requirements in the part 53
proposed rule is an explicit requirement to consider and address the
potential impact of a large, commercial aircraft, as is currently
required of parts 50 and 52 applicants under Sec. 50.150, ``Aircraft
impact assessment.'' When the Commission promulgated the aircraft
impact final rule on June 12, 2009 (74 FR 28112), it noted that ``the
impact of a large aircraft on the nuclear power plant is regarded as a
beyond-design-basis event'' and it was ``the NRC's view that effective
mitigation of the effects of events causing large fires and explosions
(including the impact of a large, commercial aircraft) can be provided
through operational actions,'' which were covered by other
requirements. In light of this view, the Commission stated that ``the
mitigation of the effects of aircraft impacts through design should be
regarded as a safety enhancement which is not necessary for adequate
protection.'' In the Regulatory Analysis that accompanied the aircraft
impact rule, the NRC quantified the costs of the rule, but did not
quantify the benefits of the rule, stating that the ``benefits of the
final rule can be evaluated only on a qualitative basis.'' The NRC
concluded that the key benefit of the rule was ``improvement in
knowledge.'' The Commission acknowledged that ``it is difficult to
quantify the safety enhancement gained through implementation of the
aircraft impact rule,'' but stated that ``the NRC nevertheless believes
that the cost of performing the assessment and incorporating the
results into the design . . . is justified in view of the increased
safety provided by implementation of the aircraft impact rule.''
It has been over 15 years since the promulgation of the aircraft
impact rule in 2009. Events like the terrorist attacks of September 11,
2001, are now much less likely due to significant increases in security
at commercial aviation facilities as well as hardened access to
aircraft cockpits. In addition, it is not clear that the Commission's
previous belief that the cost of implementation of the aircraft impact
rule was justified by the increase in safety provided by the rule would
hold true for future reactors licensed under part 53. As stated
previously, the NRC concluded that the key benefit of the rule was
``improvement in knowledge'' achieved by performing the aircraft impact
assessment. It's worth noting that licenses issued under parts 50 and
52 were largely based on deterministic analyses of the safety of the
facility relying on the General Design Criteria. The technical
requirements in part 50 were supplemented over the years to address
specific beyond-design-basis events, such as the loss of large areas of
the plant due to fires and explosions. In contrast, under part 53,
applicants will be required to perform a comprehensive assessment of
their reactor design to identify potential failures, susceptibility to
internal and external hazards, and other contributing factors that
could pose a risk to public health and safety. The spectrum of events
and hazards considered will include those that have traditionally been
considered design-basis events and those that have been considered
beyond-design-basis events. Although part 53 does not include
prescriptive requirements to assess a licensing-basis event comprising
an intentional act that could cause large fires or explosions, it does
require applicants to assess a full spectrum of unplanned events, to
include anticipated events, unlikely events, and very unlikely events.
The NRC believes that the systematic evaluations of internal hazards,
external hazards, and security threats under part 53 and part 73
sufficiently address the potential loss of large areas of the plant due
to explosions or fire currently addressed under Sec. 50.155(b)(2).
Therefore, part 53 applicants will have considered how to mitigate
the broader potential plant impacts that may result from an event such
as the impact of a large aircraft. As a result, applicants and
licensees under part 53 will have substantially more information about
the design of their facilities than applicants and licensees did before
the promulgation of the aircraft impact rule. Accordingly, the
``improvement in knowledge'' to be gained by requiring a separate
assessment of the impact of a large commercial aircraft under part 53
is expected to be significantly less than the improvements in knowledge
for part 50 or 52 applicants the Commission estimated when it
promulgated the aircraft impact rule. Because the potential impact of
beyond-design-basis events are considered in other ways under part 53,
the NRC concludes that the cost of performing a separate aircraft
impact assessment and incorporating the results into the design of a
commercial nuclear plant licensed under part 53 would not be justified.
For these reasons, this final rule does not contain requirements for
applicants to assess the impact of a large, commercial aircraft on the
design of the facility.
Section 53.450 establishes analysis requirements and centers upon
the use of a PRA, other SREs, or a combination thereof with other
generally accepted approaches for systematically evaluating engineered
systems. The use of PRA, other SREs, or a combination thereof as a key
component in the analysis requirements for part 53 reflects the decades
of improvements in the use of such methodologies and their increasing
use 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. This policy statement also acknowledges the variability in
the characteristics of events considered and the associated complexity
of engineered systems related to different regulatory activities and
that risk-informed analysis techniques of varying complexity may be
employed to yield meaningful insights and results. In that regard, the
use of PRA, other SREs, or a combination thereof under part 53 needs to
be commensurate with the complexity of the analyzed systems and their
behaviors, with consideration of all aspects of operations. 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 maintains 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, other SREs, or a combination thereof will be used together
with other techniques in part 53 to identify and categorize LBEs,
classify SSCs, evaluate defense in depth, and inform the appropriate
special treatments for SSCs. This increased role for PRA and SREs

[[Page 15708]]

necessitates that they be developed, performed, and maintained in
accordance with NRC approved standards and practices (see Sec.
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 must
be qualified for the range of conditions being simulated across a wide
range of unplanned events. These analyses must use realistic approaches
and address uncertainties associated with states of knowledge,
modeling, and performance of SSCs.
While industry consensus PRA standards and PRA peer review
processes endorsed in RGs 1.200 and 1.247 remain 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 with SREs and the 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 regulatory requirements for the specific uses of analyses under
Sec. 53.450(b). The specific uses of analyses in Sec. 53.450(b) are
to inform LBE selection; inform classification of SSCs according to
safety significance; evaluate adequacy of defense in depth; identify
and assess all plant operating states with a potential for uncontrolled
release of radioactivity to the environment; 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; and inform the establishment and updating of appropriate
measures for plant operations, including availability controls, to
ensure configurations and special treatments for SR SSCs and NSRSS SSCs
provide the capabilities, availability, and reliability consistent with
satisfying the high-level safety criteria in Sec. 53.220.
Likewise, NRC determinations of the acceptability of such PRAs
would include consideration of the appropriateness of the applicant-
defined 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.
For risk contributors that are excluded from PRA logic models or
PRA screening processes and are otherwise analyzed by an SRE--also
referred to as supplementary analyses--the NRC plans to develop
guidance for determining the acceptability of such SREs.
Section 53.450(c) requires periodic maintenance and upgrading of
the PRA, other SREs, or a combination thereof 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 Sec. 53.220 and the evaluation criteria
of Sec. 53.450(e)(2). The periodic maintenance of the PRA, other SREs,
or a combination thereof is also 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 periodic assessments performed by licensees to support the
maintenance of the PRA, other SREs, or a combination thereof and other
requirements in part 53 will 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.
Section 53.450(d) provides requirements for the qualification of
the analytical codes used in modeling the physical behavior of plant
systems and that those codes must be qualified for the range of
conditions for which they are to be used.
The categories of LBEs used in part 53 include anticipated event
sequences, unlikely event sequences, and very unlikely event sequences.
The unlikely event sequences 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 Sec. 53.450(e), the analyses of
LBEs other than DBAs will be used not only to show the performance
criteria of Sec. 53.220 are satisfied but also to show that evaluation
criteria defined for each LBE or category of LBEs are satisfied. Such
evaluation criteria for specific LBEs or categories of LBEs are 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. An applicant's or licensee's
defining of evaluation criteria under Sec. 53.450(e) and the risk
performance objectives under Sec. 53.220(b) are also part of the
integrated approach within part 53 where the analyses from subpart C
are used for decisions on design, siting, and operations. As an
example, an applicant or licensee could propose to justify siting
proposals by defining their evaluation criteria such that the
calculated consequences for an individual at the exclusion area
boundary are less than the total effective dose equivalent (TEDE)
values used in graded approaches to assessing population densities
under subpart D. Another requirement for the Sec. 53.450(e) analyses
is that the methodology must 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 maintains an important role for a deterministic analysis of
DBAs in the performance criteria of Sec. 53.210 and the related
analytical requirements in Sec. 53.450(f). The analysis of DBAs will
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 set forth in this section, DBAs must 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 Sec. Sec. 53.760 through
53.795.
While the DBAs analyzed under part 53 are similar to the
traditional DBAs analyzed under parts 50 and 52, there are important
distinctions between the overall role of DBA analyses in part 50 and
part 53. In part 53, the role of the

[[Page 15709]]

DBA analysis is 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 Sec. 53.210. The
overall control of risks posed by commercial nuclear plants under part
53 is provided by the analyses of and measures taken for both DBAs and
other LBEs, including very unlikely event sequences. This contrasts
with the traditional deterministic approach in part 50 wherein the
analyses of DBEs such as DBAs were used to provide bounding
assessments, to 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(g) includes provisions to ensure that analyses are
performed to support the design requirements of Sec. 53.440(e) on fire
protection and Sec. 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 analysis requirements
related to fire protection 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 establishes criteria for the safety classification
of SSCs and determination of appropriate special treatments. As noted
in subpart A, the term ``Special treatments'' is defined to mean those
items, such as measures taken to satisfy functional design criteria,
quality assurance, and programmatic controls, that provide assurance
that certain SSCs will provide defense in depth or perform risk-
significant functions. These requirements also provide confidence that
the SSCs will perform under the service conditions and with the
reliability credited in the analysis performed in accordance with Sec.
53.450 to satisfy the safety criteria in Sec. Sec. 53.210 and 53.220.
The terminology used in part 53 includes the following categories for
SSC classification: (1) SR; (2) NSRSS; and (3) non-safety significant.
Requirements for SR SSCs are defined in other sections of part 53 and
include using TSs for controls during operation and the application of
quality assurance requirements from appendix B to part 50.
Requirements for NSRSS SSCs include the need to identify necessary
special treatments such as performance measures on reliability.
Licensees will generally be afforded flexibility in maintaining and
changing special treatments for SSCs categorized as NSRSS. Non-safety-
significant SSCs will be addressed under normal licensee programs for
commercial grade equipment and typical industry practices for general
plant design and maintenance. Safety-related SSCs 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 Sec. 53.220 and the evaluation
criteria in Sec. 53.450(e).
Section 53.480 establishes seismic design considerations. This
section relates to the safety criteria in subpart B, the analytical
requirements related to external hazards in Sec. 53.450, and subpart
D, ``Siting Requirements.'' For licenses issued under part 53, this
section in subpart C will 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 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 Sec. 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 Sec. 53.450 must address seismic hazards as well as other
external hazards. The expected responses of SSCs to a range of seismic
events must be included in the analyses when ensuring that the safety
criteria defined under Sec. 53.220 will 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

Subpart D in part 53 states requirements for the siting of
commercial nuclear plants and serves the role provided by 10 CFR part
100, ``Reactor Site Criteria,'' for nuclear reactors licensed under
parts 50 and 52. As reflected in Sec. 53.500, the reason for
establishing siting requirements remains 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.
Section 53.510 requires 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 are used in Sec.
53.415, which requires that SR SSCs be designed to withstand the
effects of natural phenomena and constructed hazards of levels or
severities up to design-basis external hazard levels. The design-basis
levels for external hazards relevant to a site need to account for
uncertainties and variabilities in data, models, and methods used to
characterize those hazards. Existing approaches can 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 risk-informed approaches to address
seismic events are reflected in Sec. 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 Sec. 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
Performance-Based Approach to Define the Site-Specific Earthquake
Ground Motion,'' is an acceptable way to develop site-specific ground
motion

[[Page 15710]]

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 needs to be
sufficient to inform a site-specific design (e.g., a CP or custom
combined license (COL)) or confirm the use of a standard design for a
commercial nuclear plant under Sec. 53.480 and other sections of
subpart C. A risk-informed approach can use several design-basis ground
motions (DBGMs) to assess SSCs in various SDCs (i.e., one DBGM per
SDC). Section 53.510(d) states 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 requires applicants to identify and assess site
characteristics related to topics that include meteorology, geology,
hydrology, or other areas in the design and analyses required under
subpart C.
Section 53.530 sets requirements for population-related
considerations and largely maintains requirements and definitions
similar to those currently in part 100 for an exclusion area, low
population zone, 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 dose guidelines used in Sec. 53.210, which are the same as
those in the existing regulations in parts 50, 52, and 100. The
definitions in Sec. 53.020 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 long-standing preference
for siting reactors in areas of low population density is maintained in
part 53 by using the current language from part 100 as one option under
Sec. 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.'' The NRC
recognizes that safety, environmental, economic, or other factors may
justify siting commercial nuclear plants in areas with higher
population densities or within a densely populated center containing
more than about 25,000 residents. Therefore, an option is included
within Sec. 53.530 for such sites to be proposed using assessments of
additional societal risks associated with siting a reactor in areas of
higher population density (e.g., potential increases in population dose
or economic consequences from reactor accidents) in comparison to the
societal benefits of a specific site (e.g., ability to use existing
infrastructure for a retired fossil fuel power plant). Site-related
requirements in part 20 (restricted area) and part 73 (protected and
owner-controlled areas) remain applicable to commercial nuclear plants
licensed under part 53.
Section 53.540 requires that site characteristics be appropriately
considered in other activities such as the design and analysis
performed under subpart D of part 53 and the emergency planning and
security programs under subpart F of part 53.

Subpart E--Construction and Manufacturing Requirements

The part 53 language establishes construction and manufacturing
requirements in subpart E. The language for construction-related
activities largely reflects current requirements in part 50 without any
fundamental changes. Limited changes were 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 language
for requirements for manufacturing activities largely mirrors those for
construction-related activities. However, the manufacturing
requirements have been updated from 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 is not addressed by these subparts.
Section 53.600 establishes the overall construction and
manufacturing requirements for CPs, OLs, COLs, MLs, and limited work
authorizations (LWAs). This section connects the construction and
manufacturing requirements to the safety criteria, quality assurance
requirements, and other requirements located in other subparts. These
requirements require that construction and manufacturing activities be
managed and conducted such that when combined with associated design
features and programmatic controls, the constructed plant will satisfy
the relevant requirements in subpart B.
Section 53.605 establishes requirements for the reporting of
defects and instances of noncompliance during construction. This
section provides equivalent requirements to those in Sec. 50.55(e).
Section 53.610(a) establishes the requirement to have in place a
well-defined command and control structure to manage construction
activities. The requirements generally reflect current requirements,
with an emphasis on the quality assurance programs for complying with
the requirements in appendix B to part 50. Section 53.610(a)(6)
requires programmatic controls for implementing special treatment for
NSRSS SSCs to align with requirements in other subparts in part 53. The
section also refers to other NRC regulations to address matters such as
requirements to have an 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) provides requirements governing construction
activities, including the equivalent of the requirement in Sec.
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) requires 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
construction-related activities such as backfilling of excavated
portions of the site appropriately address potential pre-construction
activities such as the emplacement of retaining walls or drainage
systems. Other requirements in these paragraphs are 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) addresses 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.

[[Page 15711]]

Section 53.620(a) includes 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.
---------------------------------------------------------------------------

\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.
---------------------------------------------------------------------------

The part 53 sections in subpart E for manufacturing and in subpart
H for licensing matters maintain requirements largely 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 reactors onsite by a variety of
contractors and subcontractors.
Subpart E includes requirements that 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, Sec. 53.620 establishes 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 includes 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 requirements in Sec. 53.620(b) that 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 is also incorporated
into this 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) provides 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, subpart E refers 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 included in areas such as having a fire protection
program, an emergency plan, training programs, and procedures to
minimize contamination.
The most significant change for MLs in part 53 as compared to MLs
under part 52 relates to Sec. 53.620(d) in subpart E and the
associated licensing provisions in subpart H. These provisions 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 or OL.
The first requirement in Sec. 53.620(d) establishes limitations on
when a license under part 70 would authorize the loading of fuel into a
reactor manufactured under an ML. The regulation requires the
manufactured reactor to be configured during its loading, storage, and
transport with features to prevent criticality and that those features
be specified in the ML. The requirement provides flexibility because of
the potential variety of reactor designs, the variety of possible
measures to prevent criticality, and the range of possible conditions
associated with the loading, storage, and transport of manufactured
reactors. For example, the features to prevent criticality that could
be considered individually and collectively to address possible adverse
conditions include the reactivity control systems in place to support
operations, inherent features of the fuel and materials within a
manufactured reactor, and temporary measures or physical mechanisms
(e.g., neutron poisons) for specific circumstances and conditions, such
as during transport. This requirement contributes to the NRC's
longstanding practice of requiring defense in depth for preventing
accidents in any facility dealing with SNM, including requirements in
Sec. 70.64 for certain part 70 licensees to adhere to the ``double
contingency principle.''
The requirements to have in place features 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 features to prevent criticality in the part 53
requirements will reasonably ensure that a manufactured reactor does
not become critical over a range of possible conditions. With the
requirements for features to prevent criticality under part 53 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, these measures will reasonably ensure that the
manufactured reactor is not capable of operations, thereby obviating
the need for a COL under Sec. Sec. 53.1416 and 53.1440 to authorize
fuel loading. Additionally, this approach focuses 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, will be regulated primarily under the
part 70 license. The reference to the requirements in subpart H of part
70 in Sec. 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 10 CFR part 51, ``Environmental
Protection Regulations for Domestic Licensing and Related Regulatory
Functions,'' provide a flexible approach for environmental review to
address the range of regulated activities under part 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.
Section 53.620(d) cites the requirements in parts 70, 71, and 73 to
ensure important features and programs

[[Page 15712]]

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) includes 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 Sec.
73.67, ``Licensee fixed site and in-transit 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 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, Sec. 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 physical security program for
fueled manufactured reactors requires 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 requirements 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 necessary for physical
security, radiation monitoring, or criticality prevention.
The regulations in part 53 covering the activities related to the
storage, movement, and loading of fresh fuel into a manufactured
reactor in the manufacturing facility likewise refer to the applicable
regulations in part 70. Section 53.620(d) also requires 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 final 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) only allows the transport or removal of a
manufactured reactor or portions of a manufactured reactor for either
(1) delivery to a domestic site for which the Commission has issued a
COL or CP authorizing the construction of a commercial nuclear plant
using a manufactured reactor under the specific ML, or (2) export in
accordance with 10 CFR part 110, ``Export and Import of Nuclear
Equipment and Material.'' This requirement is similar to the
limitations in Sec. 52.153. An additional paragraph in Sec. 53.620(e)
provides 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. As noted previously, Sec. 53.620(e) includes an additional
provision that allows a manufactured reactor or portions of a
manufactured reactor to be removed from the place of manufacture for
export in accordance with part 110, which represents another difference
from the similar provision in Sec. 52.153.
Section 53.620(f) includes requirements for the acceptance and
installation of a manufactured reactor at the site of a commercial
nuclear plant. The requirements reference the construction requirements
in Sec. 53.610 to govern the integration of the manufactured reactor
into the construction of a commercial nuclear plant. Other requirements
in the section address required receipt inspections and verification
that interface requirements between the manufactured reactor and the
balance of the commercial nuclear plant have been met.

Subpart F--Requirements for Operation

Subpart F provides 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 provides the
general organization and overall objectives of subpart F, which are to
establish requirements during operations for (1) plant SSCs; (2)
personnel; and (3) plant programs.
Section 53.710 provides 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 subparts B and C. The basic
structure of this section is that measures for SR SSCs are provided by
TS and measures for NSRSS SSCs are required to be addressed with
licensee-controlled documents and procedures.
The general content and control of TS under part 53 are similar to
the requirements in part 50. The requirements for TS 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 requirements for TS
also include a section on important design elements, which is similar
to design features in Sec. 50.36, and a section for administrative
controls. A provision addressing the development and submittal of TS to
address decommissioning activities is also included in subpart G.
The requirements for TS under part 53 do not carry over safety
limits or associated limiting safety system settings from Sec. 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 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 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 in Sec. 53.410 and TS under
Sec. 53.710(a) define similar requirements as those provided by the
safety limit and limiting safety system setting requirements in Sec.
50.36.

[[Page 15713]]

The requirements for TS under part 53 do not include specific
criteria for identifying when LCOs must be established (i.e., do not
include an equivalent to Sec. 50.36(c)(2)(ii)). Instead, consistent
with subparts B and C, the TS requirements in subpart F of part 53
define TS LCOs as providing limits on SR SSCs. The SR SSCs protect
against DBAs to demonstrate compliance with the safety criteria in
Sec. 53.210. In the construct for part 53, risk-significant SSCs are
addressed through a combination of TS for SR SSCs and establishment and
monitoring of performance standards for NSRSS SSCs.
In addition to addressing TS for SR SSCs, Sec. 53.710 requires
appropriate control measures be developed and implemented for NSRSS
SSCs. Examples include appropriate surveillances and controls
established through reliability assurance programs. Configuration
management and other special treatments 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
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 Sec. 53.220 and the evaluation criteria in Sec.
53.450(e). The performance objectives for NSRSS SSCs reflect that the
comprehensive risk metrics and related risk performance objectives
established under Sec. 53.220 may involve assessing and averaging the
risks over a defined period (e.g., plant year) and do not constitute a
real-time requirement that must be continuously demonstrated by the
licensee. The controls under Sec. 53.710(b) justify 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 can 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 subpart F.
Section 53.715 provides 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 requirements are similar to those
included in Sec. 50.65 (maintenance rule), including the need to
assess and manage the increase in risk that may result from the
maintenance activities. While, for the maintenance rule, specific
criteria must be developed to capture both SR and non-SR but otherwise
important SSCs, Sec. 53.715 covers SR SSCs and NSRSS consistent with
other subparts in part 53.
Section 53.720 provides the requirements for responding to a
seismic event during the operating phase of the life cycle of a
commercial nuclear plant and is equivalent to the requirements in
paragraph IV(a)(3) of appendix S, ``Earthquake Engineering Criteria for
Nuclear Power Plants,'' to part 50.
Part 53 includes 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 Advisory Committee on Reactor
Safeguards (ACRS). Key considerations include the recognition that
staffing, operator qualifications, and HFE are interconnected areas
that must be 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 are in Sec. Sec.
53.725 through 53.830. Section 53.725 discusses applicability and
defines specific terms. Some definitions draw from those in Sec. 55.4.
Several new definitions are introduced for use within the context of
subpart F. These new definitions are the following: ``Automation,''
``Auxiliary operator,'' ``Generally licensed reactor operator,''
``Interaction-dependent-mitigation facility,'' ``Load following,'' and
``Self-reliant-mitigation facility.''
Sections 53.725 through 53.830 are divided into four portions that
cover general operational requirements, operator and senior operator
licensing requirements, 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 will be required to
use GLROs upon demonstrating compliance with the criteria in Sec.
53.800.
Certain routine communications are necessary to facilitate the
operator licensing process. The NRC adapts the requirements of
Sec. Sec. 55.5 and 50.74 to Sec. 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 adapts the
requirements of Sec. 55.8, to include any needed updates in OMB
approval information, to Sec. 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 adapts the requirements of
Sec. 55.9 to Sec. 53.728 to require the completeness and accuracy of
material information provided by individual applicants and license
holders.
Section 53.730 provides performance-based and technology-inclusive
requirements for assessing the role of personnel in facility safety,
applying human system considerations within facility design, and
incorporating operational approaches that are consistent with design-
specific safety considerations. Most of these requirements are adapted
from portions of Sec. Sec. 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
intends that these technical requirements will, together, serve as a
component of the required content of applications for OLs and COLs
under part 53. Additionally, the NRC intends that the specific
technical requirements associated with HFE, human-system interface
design, concept of operations, functional requirements analysis, and
function allocation will 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

[[Page 15714]]

NRC adapts Sec. 53.730(a) from the HFE design requirements of Sec.
50.34(f)(2)(iii). A key difference is that the requirement is now
focused on settings where personnel fulfill their safety or emergency
response roles wherever they may occur. The NRC additionally includes
within the scope of this requirement activities for assuring the
continued availability of plant equipment that is needed for safety,
and the NRC envisions that these activities may encompass relevant
maintenance, inspections, and testing as well. This requirement is
associated with the staff guidance for conducting scalable reviews of
HFE in DRO-ISG-2023-03, ``Development of Scalable HFE Review Plans''
that accompanies 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 requires under Sec. 53.730(b) such human-system
interface design features for all facilities, irrespective of other
flexibilities under part 53. Therefore, the NRC adapts specific post-
Three Mile Island requirements of Sec. 50.34(f) in a technology-
inclusive manner as detailed in the following:
<bullet> Paragraph (b)(1) is adapted from Sec. 50.34(f)(2)(iv).
<bullet> Paragraph (b)(2) is adapted from Sec. 50.34(f)(2)(v).
<bullet> Paragraph (b)(3) is adapted from Sec. 50.34(f)(2)(xi),
50.34(f)(2)(xii), and 50.34(f)(2)(xxi).
<bullet> Paragraph (b)(4) is adapted from Sec. 50.34(f)(2)(xvii),
50.34(f)(2)(xviii), 50.34(f)(2)(xix), and 50.34(f)(2)(xxiv).
<bullet> Paragraph (b)(5) is adapted from Sec. 50.34(f)(2)(xxvi).
<bullet> Paragraph (b)(6) is adapted from Sec. 50.34(f)(2)(xxvii).
In addition to the requirements of Sec. 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 is provided
under Sec. 53.730(b)(7). This prescriptive set of design requirements
for those facilities that demonstrate compliance with the criteria of
Sec. 53.800 recognizes that the application of HFE under Sec.
53.730(a) is anticipated to be significantly streamlined 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 is
conservatively mandated irrespective of any other design considerations
for both interaction-dependent and self-reliant mitigation facilities,
as required under Sec. 53.730(b)(8).
The NRC requires under Sec. 53.730(c) 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 requires under Sec. 53.730(d) 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 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 this final rule is the further
need not only to 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 adapts in Sec. 53.730(e) the requirements of
Sec. 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 Sec. 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 requires under Sec.
53.730(f) 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
staffing plan description of how engineering expertise will be provided
should include details of the position, such as location, expected
response time, access to plant status information, and methods of
communication. The staffing plan description should contain information
on how the described response time has been or will be determined to be
adequate based on the facility design. This requirement is associated
with the staff guidance for reviewing operations staffing plans in 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'' that
accompanies part 53. Following NRC approval of the OL or COL, the
staffing plan will become a condition of the facility license.
The NRC intends that, at a minimum, the approved licensed 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
therefore be required for any subsequent changes.
Operator training and qualification programs provide an essential

[[Page 15715]]

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, under part 53 the NRC allows facilities to develop
operator proficiency programs based on facility-specific
considerations.
Therefore, the NRC requires in Sec. 53.730(g)(1), as part of its
approval of the OL or COL, approval 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 requires in Sec.
53.730(g)(2) 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 will 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 adapts the requirements of Sec. 50.54(i-1) in
Sec. 53.730(g)(3) to require that operator continuing training
programs be in effect to support operator performance. Under part 53,
the NRC requires 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 Sec. 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 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 adapts
the requirements of Sec. 55.13 in Sec. 53.735 to accomplish this,
with a notable difference being the incorporation of GLROs.
Section 53.740 provides requirements for OL and COL holders under
part 53. Portions of Sec. 53.740 are adapted from the conditions of
Sec. 50.54. In general, the conditions for operations staffing under
part 53 reflect considerations for potential technological differences
and varying concepts of operation that are expected among part 53
facility licensees. Additionally, certain requirements are specific to
the operating phase while others remain in effect following the
permanent cessation of facility operations during the decommissioning
phase.
All commercial nuclear plants licensed under part 53 require some
form of licensed operator staffing, whether it be by specifically or
generally licensed operators. Consistent with this, the NRC requires
under Sec. 53.740(a) that facility licensees 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 technology-inclusive 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 requires
under Sec. 53.740(b) 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
Sec. 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 adapts the requirements of Sec. 50.54(i) in Sec.
53.740(c) to require that only specifically licensed operators and
senior operators or, alternatively, 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 adapts the requirements of Sec. 50.54(l) in
Sec. 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 therefore adapts the
requirements of Sec. 50.54(j) in Sec. 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 Sec. 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. 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) allows 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

[[Page 15716]]

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 adapts the
requirements of Sec. 50.54(m)(2)(iv) in Sec. 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
will already be required by either a specifically licensed senior
operator or by a GLRO, the NRC omits 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 adapts the requirements of Sec. 50.54(x) and (y) in Sec.
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 extends this authority to GLROs. While it is not anticipated that
GLROs will have a role in the fulfillment of safety functions at self-
reliant-mitigation facilities, nor is it anticipated that operators at
such facilities would be in a position by which to significantly
influence radiological safety outcomes, the very nature of the Sec.
50.54(x) and (y) and Sec. 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
AEA authorizes the Commission to prescribe conditions for the licensing
of operators and to issue licenses consistent with those conditions.
The NRC adapts the requirements of Sec. 55.3 in Sec. 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 will license individuals as operators under both specific
and general licensing frameworks. Specific licenses will be for
licensed operators (i.e., reactor operators) and senior operators
(i.e., senior reactor operators) and will be issued to a named person
upon approval by the Commission of an application for that named
person. In contrast, GLROs will perform duties under the provisions of
a general license that is effective without the filing of an
application with the Commission or the issuance of licensing documents
to a particular person. The NRC sets forth requirements for the use of
a specific licensing process for licensed operators and senior
operators under Sec. Sec. 53.760 through 53.795, with Sec. 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 public health and safety. In addition to a requirement for
medical fitness, a medical examination by a physician to confirm
compliance with this requirement is necessary. The NRC adapts the
requirements of Sec. Sec. 55.21, 55.23, and 55.27 under Sec. 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 does not extend a comparable medical requirement
to GLROs.
The NRC also adapts the requirements of Sec. Sec. 55.25 and
50.74(c) in Sec. 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 adapts
the requirements of Sec. Sec. 55.31 through 55.35 in Sec. 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 provides additional flexibility by locating
certain details associated with the preparation and submittal of these
applications within guidance in lieu of placement within this final
rule itself.
The NRC includes overall programmatic requirements for specifically
licensed operator and senior operator training, examination, and
proficiency in Sec. 53.780. In general, the requirements are adapted
from those in part 55, with several additional flexibilities being
incorporated to better account for potential variations in reactor
technologies and concepts of operations. The requirements in Sec.
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

[[Page 15717]]

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 requires under Sec. 53.780(a) that facility
licensees 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 requires
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 adapts the requirements of
Sec. Sec. 55.40, 55.41, 55.43, and 55.45 in Sec. 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
is 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)
must provide for acceptable levels of both test validity and test
reliability in order to be considered acceptable. The NRC intends that
staff guidance will be available to facilitate the review of licensing
examination programs that are proposed by facility licensees and that,
following NRC approval, initial examination programs will be subject to
an appropriate change control process. Furthermore, the NRC provides
holders of licenses to operate commercial nuclear plants under part 53
the alternative of administering their own approved licensing
examinations. The NRC will 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:

<bullet> Reactor Theory, Thermodynamics, and Chemical Interactions
<bullet> Plant Systems and Components
<bullet> Reactivity Management and Manipulations
<bullet> Radiation Control and Safety
<bullet> Emergency, Abnormal, and Normal Operations
<bullet> 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 adapts the requirements of Sec. 55.59 in Sec.
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 is that distinct
annual operating test and biennial written examination components are
not mandated, with the facility licensee instead proposing the
examination methods and criteria to be used in assessing satisfactory
performance. The NRC intends that guidance will 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 will be subject to an appropriate
change control process.
For examinations to provide valid assessments of the knowledge and
abilities of individuals, the examinations must remain free from
compromises that could affect their underlying integrity. The NRC
adapts the requirements of Sec. 55.49 in Sec. 53.780(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
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 adapts the requirements of Sec. 55.46 in Sec.
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
requirements of part 53 do not mandate that full scope, plant-
referenced simulators be used and will 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
adapts the requirements of Sec. 55.47 in Sec. 53.780(f) to allow for
consideration of requests for waivers of examinations requirements. In
contrast with the part 55 requirements, the NRC locates certain details
associated with such waiver requests within guidance documentation in
lieu of placement within this final rule itself.
For licensed operators and senior operators to perform their
assigned duties safely and reliably, it is essential that they perform
those duties frequently enough so as to maintain a sufficient degree of
proficiency. The NRC adapts the requirements of Sec. 55.53(e) and (f)
in Sec. 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, in contrast with
the requirements of part 55, is allowing 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

[[Page 15718]]

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 adapts the requirements of Sec. 55.53 in Sec. 53.785 to require
appropriate conditions of licenses for specifically licensed operators
and senior operators. However, in contrast with the requirements of
Sec. 55.53(e) and (f), the NRC is allowing 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 adapts the requirements of Sec. Sec. 55.51 and
55.61 in Sec. 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 6 years, after which they expire,
unless otherwise renewed. The NRC adapts the requirements of Sec. Sec.
55.55 and 55.57 in Sec. 53.795 to address the expiration and renewal
of licenses issued to specifically licensed operators and senior
operators.
In developing this final 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 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 establishes under Sec. 53.800 a new
class of facility (defined as a self-reliant-mitigation facility),
according to the criteria contained in Sec. 53.800 for part 53. These
facilities will 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 Sec. 53.800 will instead be considered interaction-
dependent-mitigation facilities and will 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 differ from specifically
licensed operators because the latter will 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 in Sec. 53.800 designate self-reliant-mitigation
facilities. These criteria are derived from the following set of
considerations:

<bullet> no human action needed to satisfy radiological consequence
criteria;
<bullet> no human action needed to address LBEs;
<bullet> safety functions not allocated to human action;
<bullet> reliance upon robust and highly reliable safety features; and
<bullet> appropriate defense in depth achieved without reliance on
important human action.

It should be noted that those facilities not meeting the criteria in
Sec. 53.800 will instead be classified as interaction-dependent-
mitigation facilities and will require staffing by specifically
licensed operators and senior operators instead.
Generally licensed reactor operators will perform duties under the
provisions of a general license that is effective without the filing of
an application with the Commission or the issuance of licensing
documents to a particular person. The NRC sets forth requirements for
the general licensing process for GLROs under Sec. Sec. 53.805 through
53.820. The requirements for GLROs parallel those for senior operators
in regard to their comparable administrative responsibilities.
Nonetheless, the requirements for GLROs are 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-reliant-mitigation facilities.
In order to use GLROs in lieu of specifically licensed operators
and senior operators, a OL/COL applicant must demonstrate that its
proposed facility is a self-reliant-mitigation 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
are 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 Sec. 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 are
predicated on this assumption. The NRC therefore establishes under
Sec. 53.805 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 Sec. 53.800 after the
issuance of an OL or COL will constitute a reportable event (i.e., an
unanalyzed condition that significantly degrades plant safety) under
the provisions of Sec. 53.1630.
The NRC sets forth the general license for GLROs under Sec.
53.810. GLROs will be licensed as a class of individuals under the
provision of Sec. 53.810(a) and will be subject to the conditions
specified in Sec. 53.810(b) through (g). Portions of these conditions
are adapted from Sec. 55.53 and from those conditions currently
included in the licenses issued to specifically licensed operators and
senior operators. The NRC retains the ability to suspend or prohibit
individuals from operating under the general license should such action
be warranted.
The NRC includes overall programmatic requirements for GLRO
training, examination, and proficiency under Sec. 53.815. In general,
these

[[Page 15719]]

requirements are adapted from those of part 55 and parallel those also
included for specifically licensed senior operators in Sec. 53.780.
These requirements include increased flexibilities and several targeted
relaxations that reflect the limited role of GLROs in facility safety.
The requirements under Sec. 53.815 cover, in part, the initial
training, initial examination, continuing training, requalification
examination, and proficiency of GLROs. Section 53.805 requires the
facility licensee to develop, implement, and maintain these programs.
Section 53.810, in turn, prescribes that the requirements of Sec.
53.805 must be met as a requirement of the general license. The
implication of this structure is that the facility licensee must
implement these programs for training, examination, and proficiency,
and GLROs must 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 an 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 requires under
Sec. 53.815(b) that facility licensees implement a SAT-based 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 requires that such programs be
subject to NRC approval, oversight, and appropriate change control
processes. The training program must ensure that 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 adapts the requirements of Sec. Sec.
55.40, 55.41, 55.43, and 55.45 in Sec. 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
is that facility licensees are 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)
must provide for acceptable levels of both test validity and test
reliability in order to be considered acceptable. The NRC intends that
staff guidance will be available to facilitate the review of initial
examination programs that are proposed by facility licensees and that
approved initial examination programs will be subject to an appropriate
change control process. In contrast with both the requirements of part
55 and the requirements of Sec. 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:

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 adapts the requirements of Sec. 55.59 in Sec.
53.815(b) to require that facility licensees implement both an SAT-
based continuing training program and a requalification examination
program. However, a notable difference from the examinations required
under part 55 is that distinct annual operating test and biennial
written examination components are not mandated. The facility licensee
will instead propose examination methods and criteria to be used in
assessing satisfactory performance. Furthermore, unlike the comparable
requirements of part 55 and those for specifically licensed operators
and senior operators, a biennial periodicity for requalification
examinations is not prescribed. However, adequate justification for the
proposed periodicity of requalification examinations is required. The
NRC intends that staff guidance will be available to facilitate the
review of the requalification examination programs that are proposed by
facility licensees. Approved requalification examination programs will
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
adapts the requirements of Sec. 55.49 in Sec. 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 adapts the requirements of
Sec. 55.46 in Sec. 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 is not mandated. The potential
use of alternative simulation facilities consisting of, for example,
partial scope simulators or the plant itself, is allowed provided that
all associated requirements are 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 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, under Sec. 53.815(f) the NRC allows 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.

[[Page 15720]]

However, the NRC recognizes that facilities that utilize GLROs may have
concepts of operation that warrant unique proficiency considerations.
Therefore, the NRC requires in Sec. 53.815(g) 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 requires under Sec. 53.820 that the general license ceases 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 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 Sec.
50.120, ``Training and qualification of nuclear power plant
personnel.'' The NRC adapts under Sec. 53.830, with modifications, the
requirements of Sec. 50.120 for use in part 53 to provide more
flexible personnel training and qualification requirements than those
in Sec. 50.120 and better reflect diverse concepts of operations.
The NRC recognizes that the categories of nuclear power plant
personnel in Sec. 50.120 may not be needed for the diverse concepts of
operations, staffing models, and non-traditional personnel roles and
responsibilities anticipated under part 53; conversely, and for the
same reasons, additional categories of personnel may need to be covered
by part 53. The NRC also recognizes that the timeframe prescribed in
Sec. 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 Sec. 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 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 requires under Sec. 53.830 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 will cover the training and qualification
of personnel in the general categories of supervisors, technicians, and
other appropriate operating personnel. Regarding the category of
supervisors, this is intended to reflect on-shift supervisors for the
licensed operators, similar to the current classification in Sec.
50.120(b)(2)(iii), but Sec. 53.830 uses language that is less specific
to account for different conduct of operations and organizational
structures for commercial nuclear plants which may require greater
regulatory flexibility. The licensee is not 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 accreditation process.
Section 53.845 requires 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 programmatic requirements in subpart
F also address areas such as radiation protection needed to control
routine effluents during normal operations. Sections 53.850 through
53.910 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
is required to assess whether additional programs are needed for the
specific reactor design and location of the commercial nuclear plant.
Licensees are able to combine, separate, and otherwise organize
programs and related documents as appropriate for the technologies and
organizations associated with the commercial nuclear plant.
Section 53.850 requires 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
requirements related to doses from normal operations, including routine
effluents, are similar to those specified in Sec. 50.36a, ``Technical
specifications on effluents from nuclear power reactors,'' and related
requirements in standard TS for offsite dose calculation manuals. While
the section includes requirements that are technically and
programmatically similar to part 50, Sec. 53.850 does not include a
requirement for effluent-related TS as is required in Sec. 50.36a. A
requirement similar to that found in the administrative controls
section of TS for operating reactors licensed under parts 50 and 52 is
included for programmatic controls of solid wastes to complement the
design requirements in Sec. 53.425.
Section 53.855 requires an emergency response plan that
demonstrates compliance with the requirements in appendix E to part 50
and Sec. 50.47(b) or Sec. 50.160. The regulations in Sec. 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
complying with the applicable standards set forth in either Sec.
50.160 or the requirements in appendix E to part 50 and Sec. 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

[[Page 15721]]

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 defense-in-depth protection to the surrounding
population.'' Consistent with this policy statement, Sec. 53.855
contributes an additional independent layer of defense in depth for
commercial nuclear plants. Therefore, the emergency plans and emergency
planning zones under Sec. 53.855 are not used to demonstrate
compliance with subpart B and subpart C of part 53. Rather, compliance
with the requirements in Sec. 53.855 provides reasonable assurance
that adequate protective measures can and will be taken to protect
public health and safety in the event of a radiological emergency.
Section 53.860 identifies 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 IV, ``Changes to Other Parts of 10
CFR,'' of this document.
Section 53.860(a) requires licensees to develop, implement, and
maintain a physical protection program that meets either Sec. 73.55 or
Sec. 73.100, and includes physical protection of SNM and Category 1
and Category 2 radioactive material, if applicable.
Section 53.860(b) requires licensees to establish, implement, and
maintain an FFD program under part 26. Section 53.860(c) requires
licensees to establish, implement, and maintain an AA program in
accordance with either Sec. 73.56 or Sec. 73.120, as appropriate.
Section 53.860(d) requires licensees to establish, implement, and
maintain a cybersecurity program in accordance with either Sec. 73.54
or Sec. 73.110. Section 53.860(e) requires licensees to establish,
implement, and maintain an information protection system that complies
with the requirements of Sec. Sec. 73.21, 73.22, and 73.23, as
applicable.
Section 53.865 establishes requirements for quality assurance and
refers to appendix B to part 50 for the part 53 requirements for SR
design features. Requirements related to evaluating and reporting
changes to the quality assurance program are included in subpart I and
are equivalent to those found in Sec. 50.54.
Section 53.870 requires licensees to actively assess possible
degradation of SSCs from the effects of aging, fatigue, and
environmental conditions. The 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 support
performance-based approaches to the initial licensing of new
technologies. The performance-based approach to integrity assessment
programs also allows for the subsequent consideration of operating
experience and appropriate corrective actions or allowable relaxations
for ensuring that design features comply with the functional design
criteria of Sec. Sec. 53.410 and 53.420. The program is 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
can 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.
Section 53.875 establishes requirements for a fire protection
program supporting operations similar to Sec. 50.48. The fire
protection program during operations will work in concert with specific
fire protection requirements in subpart C for design and analyses and
in subpart E for construction and manufacturing.
Section 53.880 establishes 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 Sec. 50.55a. While part 53 does not incorporate
specific consensus codes and standards into the regulations, Sec.
53.880 allows for the use of generally accepted codes and standards.
The requirement for an ISI and IST program reinforces 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 requirements in part 53.
Section 53.910 establishes 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 section are similar to the requirements found
in the administrative controls section of TS for plants licensed under
parts 50 and 52. The inclusion, where appropriate, of accident
management guidelines in these requirements is intended to ensure that
an integrated set of procedures and guidelines is established by
licensees to ensure command and control across the spectrum of possible
event sequences. The required procedures also include those needed to
complement the design requirements in Sec. 53.440(m) related to
criticality alarms and the equivalent of the procedures required in
Sec. 50.54(hh) to address notifications of potential aircraft threats.

Subpart G--Decommissioning Requirements

Subpart G provides the regulatory requirements for the
decommissioning phase of the life cycle of a commercial nuclear plant.
The requirements in subpart G for the decommissioning of a commercial
nuclear plant are adapted from the current regulations in Sec. 50.75,
``Reporting and recordkeeping for decommissioning planning,'' Sec.
50.82, ``Termination of license,'' and Sec. 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
subpart G with relatively few changes, the requirem

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