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Proposed Rule2024-23434

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

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Published

October 31, 2024

Issuing agencies

Nuclear Regulatory Commission

Abstract

The U.S. Nuclear Regulatory Commission (NRC) is proposing to revise the NRC's regulations by adding a risk-informed, performance- based, and technology-inclusive regulatory framework for commercial nuclear plants in response to the Nuclear Energy Innovation and Modernization Act (NEIMA). The NRC plans to hold a public meeting to promote full understanding of the proposed rule and facilitate public comments.

Full Text

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[Federal Register Volume 89, Number 211 (Thursday, October 31, 2024)]
[Proposed Rules]
[Pages 86918-87128]
From the Federal Register Online via the Government Publishing Office [<a href="http://www.gpo.gov">www.gpo.gov</a>]
[FR Doc No: 2024-23434]

[[Page 86917]]

Vol. 89

Thursday,

No. 211

October 31, 2024

Part II

Nuclear Regulatory Commission

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10 CFR Parts 1, 2, 10, et al.

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

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

[[Page 86918]]

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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: Proposed rule.

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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is proposing to
revise the NRC's regulations by adding a risk-informed, performance-
based, and technology-inclusive regulatory framework for commercial
nuclear plants in response to the Nuclear Energy Innovation and
Modernization Act (NEIMA). The NRC plans to hold a public meeting to
promote full understanding of the proposed rule and facilitate public
comments.

DATES: Submit comments by December 30, 2024. Comments received after
this date will be considered if it is practical to do so, but the NRC
is able to ensure consideration only for comments received before this
date.

ADDRESSES: You may submit comments by any of the following methods
however, the NRC encourages electronic comment submission through the
Federal rulemaking website:
<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#d49cb1b8b1bafa97bcb5bab394baa6b7fab3bba2">[email&#160;protected]</a>. For technical questions contact the
individuals listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
<bullet> Email comments to: <a href="/cdn-cgi/l/email-protection#fcae899099919d9795929bd2bf9391919992888fbc928e9fd29b938a">[email&#160;protected]</a>. If you do
not receive an automatic email reply confirming receipt, then contact
us at 301-415-1677.
<bullet> Fax comments to: Secretary, U.S. Nuclear Regulatory
Commission at 301-415-1101.
<bullet> Mail comments to: Secretary, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, ATTN: Rulemakings and
Adjudications Staff.
<bullet> Hand deliver comments to: 11555 Rockville Pike, Rockville,
Maryland 20852, between 7:30 a.m. and 4:15 p.m. eastern time, Federal
workdays; telephone: 301-415-1677.
You can read a plain language description of this proposed rule at
<a href="https://www.regulations.gov/docket/NRC-2019-0062">https://www.regulations.gov/docket/NRC-2019-0062</a>. For additional
direction on obtaining information and submitting comments, see
``Obtaining Information and Submitting Comments'' in the SUPPLEMENTARY
INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Robert Beall, Office of Nuclear
Material Safety and Safeguards, telephone: 301-415-3874; email:
<a href="/cdn-cgi/l/email-protection#4d1f222f283f39630f282c21210d233f2e632a223b">[email&#160;protected]</a>; or Anders Gilbertson, Office of Nuclear Reactor
Regulation, telephone: 301-415-1541; email: <a href="/cdn-cgi/l/email-protection#54153a303126277a133d3836312620273b3a143a26377a333b22">[email&#160;protected]</a>.
Both are staff of the U.S. NRC, Washington, DC 20555-0001.

SUPPLEMENTARY INFORMATION:

Executive Summary

A. Need for the Regulatory Action

On January 14, 2019, the President signed the Nuclear Energy
Innovation and Modernization Act (NEIMA) into law (Pub. L. 115-439).
NEIMA section 103(a)(4) directs the NRC to ``complete a rulemaking to
establish a 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, Code of Federal Regulations (as in effect on the date of
enactment of [NEIMA])), with significant improvements compared to
commercial nuclear reactors under construction as of the date of
enactment of [NEIMA].''
The NRC initially considered establishing the scope of proposed
part 53, ``Risk-Informed, Technology-Inclusive Regulatory Framework for
Commercial Nuclear Plants,'' of title 10 of the Code of Federal
Regulations (10 CFR) as being for ``advanced nuclear plants''
consisting of one or more ``advanced nuclear reactors'' as defined in
NEIMA. Based on public discussions on the use of the term, the NRC
determined that the NEIMA definition, although broad, did not define
``significant improvements'' with enough specificity to implement in
NRC regulations. Additionally, a number of stakeholders suggested that
the descriptor, ``advanced,'' implied enhanced safety, while the NEIMA
definition includes ``significant improvements'' in areas other than
safety enhancements. In response to this feedback, and to be technology
inclusive, the NRC determined that the broader term ``commercial
nuclear plant'' would be preferable.
The current application and licensing requirements in 10 CFR part
50, ``Domestic Licensing of Production and Utilization Facilities,''
and 10 CFR part 52, ``Licenses, Certifications, and Approvals for
Nuclear Power Plants,'' were primarily developed to address license
requests concerning water-cooled reactors, and to address operational
requirements for those types of reactors. This proposed rule responds
to NEIMA by creating an alternative regulatory framework for licensing
future commercial nuclear plants. The new alternative requirements and
implementing guidance would adopt technology-inclusive approaches and
use risk-informed and performance-based techniques to ensure an
equivalent level of safety to that of operating commercial nuclear
plants while providing flexibility for licensing and regulating a
variety of technologies and designs for commercial nuclear reactors.

B. Major Provisions

Major provisions of this proposed rule, supported by accompanying
guidance, include the following:
<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 draft regulatory analysis to determine the
expected quantitative costs and benefits of this proposed rule and
associated guidance as well as qualitative factors to be considered in
the NRC's rulemaking decision. The conclusion from the

[[Page 86919]]

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

Table of Contents

I. Obtaining Information and Submitting Comments
A. Obtaining Information
B. Submitting Comments
II. Background
A. NRC Advanced Reactor Readiness
B. Stakeholder Views on Part 53 Preliminary Proposed Rule
Language
III. Discussion
A. Objective and Applicability
B. Need for Changes to the Existing Regulatory Framework
C. 10 CFR Part 53: Framework
IV. Part 53: Framework
Subpart A--General Provisions
A. Discussion of Definitions in Proposed Part 53
B. Other General Provisions
Subpart B--Technology-Inclusive Safety Requirements
Subpart C--Design and Analysis Requirements
Subpart D--Siting Requirements
Subpart E--Construction and Manufacturing Requirements
Subpart F--Requirements for Operation
Subpart G--Decommissioning Requirements
Subpart H--Licenses, Certifications, and Approvals
Subpart I--Maintaining and Revising Licensing Basis Information
Subpart J--Reporting and Other Administrative Requirements
Subpart M--Enforcement
V. Changes to Other Parts of 10 CFR Chapter I
10 CFR Part 26
A. Introduction
B. Proposed Changes to Part 26, Subparts A Through E and I
C. Proposed Requirements for Part 26, Subpart M
D. Proposed Changes to Part 26, Subpart N
E. Proposed Changes to Part 26, Subpart O
10 CFR Part 50
A. Section 50.160: Emergency Preparedness for Small Modular
Reactors, Non-Light-Water Reactors, and Non-Power Production or
Utilization Facilities
B. Appendix B to Part 50: Quality Assurance Criteria for Nuclear
Power Plants and Fuel Reprocessing Plants
10 CFR Part 73
A. Section 73.100: Technology-Inclusive Requirements for
Physical Protection of Licensed Activities at Commercial Nuclear
Plants Against Radiological Sabotage
B. Section 73.110: Technology-Inclusive Requirements for
Protection of Digital Computer and Communication Systems and
Networks
C. Section 73.120: Access Authorization Program for Commercial
Nuclear Plants
VI. Specific Requests for Comments
VII. Section-by-Section Analysis
VIII. Regulatory Flexibility Certification
IX. Regulatory Analysis
X. Backfitting and Issue Finality
XI. Cumulative Effects of Regulation
XII. Plain Writing
XIII. Environmental Assessment and Proposed Finding of No
Significant Environmental Impact
XIV. Paperwork Reduction Act
XV. Criminal Penalties
XVI. Voluntary Consensus Standards
XVII. Availability of Guidance
XVIII. Public Meeting
XIX. Availability of Documents

I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID NRC-2019-0062 when contacting the NRC
about the availability of information for this action. You may obtain
publicly available information related to this action by any of the
following methods:
<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.
<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 Web-based ADAMS
Search.'' For problems with ADAMS, please contact the NRC's Public
Document Room (PDR) reference staff at 1-800-397-4209, at 301-415-4737,
or by email to <a href="/cdn-cgi/l/email-protection#8bdbcfd9a5d9eef8e4fef9e8eecbe5f9e8a5ece4fd">[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#5e0e1a0c700c3b2d312b2c3d3b1e302c3d70393128">[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.

B. Submitting Comments

The NRC encourages electronic comment submission through the
Federal rulemaking website (<a href="https://www.regulations.gov">https://www.regulations.gov</a>). Please
include Docket ID NRC-2019-0062 in your comment submission. To
facilitate NRC review, please distinguish between comments on the
proposed rule and comments on the proposed guidance.
The NRC cautions you not to include identifying or contact
information that you do not want to be publicly disclosed in your
comment submission. The NRC will post all comment submissions at
<a href="https://www.regulations.gov">https://www.regulations.gov</a> as well as enter the comment submissions
into ADAMS. The NRC does not routinely edit comment submissions to
remove identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment into ADAMS.

II. Background

A. NRC Advanced Reactor Readiness

In its ``Policy Statement on the Regulation of Advanced Nuclear
Power Plants,'' dated July 8, 1986, the Commission stated that it
considered the term ``advanced'' to apply to reactors that are
significantly different from current (i.e., current in 1986) generation
light-water reactors (LWRs) then under construction or in operation,
and that ``advanced'' includes reactors that provide enhanced margins
of safety or utilize simplified inherent or other innovative means to
accomplish their safety functions. At the time, certain high
temperature gas-cooled reactors, liquid metal reactors, and LWRs of
innovative design were considered to be ``advanced.'' The 1986 policy
statement

[[Page 86920]]

provided the Commission's policy regarding the review of, and desired
characteristics associated with, advanced reactors. The NRC updated
this statement in the ``Policy Statement on the Regulation of Advanced
Reactors,'' dated October 14, 2008 (Advanced Reactor Policy Statement).
The agency has undertaken many activities related to advanced
reactors, including issuing an advance notice of proposed rulemaking
titled, ``Approaches to Risk-Informed and Performance-Based
Requirements for Nuclear Power Reactors,'' dated May 4, 2006 (71 FR
26267). These efforts were often done in parallel, and sometimes
interwoven, with the NRC's efforts to improve risk-informed and
performance-based approaches within the agency (e.g., the Commission's
policy statement, ``Use of Probabilistic Risk Assessment Methods in
Nuclear Regulatory Activities,'' dated August 16, 1995 (PRA Policy
Statement)).
In 2016, the NRC issued ``NRC Vision and Strategy: Safely Achieving
Effective and Efficient Non-Light-Water Mission Readiness'' (Advanced
Reactor Vision and Strategy Document), in response to increasing
interest in advanced reactor designs. The NRC considered the Department
of Energy's (DOE's) advanced reactor deployment goals in developing the
Advanced Reactor Vision and Strategy Document. Since publication of the
document, the NRC continues to manage its activities to support the
DOE's deployment goals. The Advanced Reactor Vision and Strategy
Document identified initiating and developing a new risk-informed and
performance-based regulatory framework as a possible long-term goal.
However, the NRC staff's initial efforts were focused on resolving
policy issues and developing guidance for licensing non-LWR
technologies under the existing regulatory frameworks (parts 50 and
52). The NRC staff issues annual Commission papers on the status and
progress of the NRC staff's activities related to advanced reactors
(e.g., SECY-24-0020, ``Advanced Reactor Program Status,'' dated
February 27, 2024). These Commission papers provide status updates for
advanced reactor activities undertaken both prior to and after
initiation of this rulemaking.
In 2017, the NRC staff prioritized activities to support the
development of technology-inclusive, risk-informed, and performance-
based licensing approaches that could be implemented under the existing
regulatory framework in parts 50 and 52. One key element of these
efforts was the Licensing Modernization Project (LMP), a cost-shared
initiative led by nuclear utilities and supported by DOE. The LMP is a
technology-inclusive, risk-informed, and performance-based methodology
developed for non-LWR designs. The LMP provides a systematic and
reproducible process for licensing-basis event (LBE) selection and
evaluation; classification of structures, systems, and components
(SSCs); and assessment of defense in depth. The LMP refined the DOE's
Next Generation Nuclear Plant Program methodologies to reflect
interactions with the NRC, to address feedback from industry, and to
broaden the scope of the approach to ensure applicability to various
non-LWR technologies. The LMP activities led to the publication and
submittal of Nuclear Energy Institute (NEI) 18-04, Revision 1, ``Risk-
Informed Performance-Based Technology Inclusive Guidance for 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 approach.
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 LMP and NEI-18-04 in
SECY-19-0117, ``Technology-Inclusive, Risk-Informed, and Performance-
Based Methodology to Inform the Licensing Basis and Content of
Applications for Licenses, Certifications, and Approvals for Non-Light-
Water Reactors,'' dated December 2, 2019. In that paper, the staff
described the relationship between the LMP and NEI-18-04 and previous
relevant Commission decisions, including those described in SECY-93-
092, ``Issues Pertaining to the Advanced Reactor (PRISM, MHTGR, and
PIUS) and CANDU 3 Designs and their Relationship to Current Regulatory
Requirements,'' dated April 8, 1993. The Commission approved the use of
the LMP methodology and NEI-18-04 as a reasonable approach for
establishing key parts of the licensing basis and content of
applications for licenses, certifications, and approvals for non-LWRs
in Staff Requirements Memorandum (SRM) SRM-SECY-19-0117, dated May 26,
2020. Although the LMP approach is technology- inclusive, the industry
and NRC staff initially focused the LMP's applicability on 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, the NRC
staff developed part 53 by building upon recent and ongoing activities
such as the LMP approach described in SECY-19-0117. Such an approach
supports implementing the NEIMA requirement to use, where appropriate,
risk-informed and performance-based techniques, and it also capitalizes
on previous initiatives by the industry, DOE, and the NRC, including
the LMP. This approach highlights the role of PRA in risk-informed and
performance-based approaches to identifying enhanced safety margins
that can be used to justify operational flexibilities. The proposed
framework is largely based on the methodology described in SECY-19-0117
and includes a prominent role for PRA.
As discussed in section II.B, ``Stakeholder Views on Part 53
Preliminary Proposed Rule Language,'' of this document, the NRC
conducted extensive public outreach on early versions of the proposed
rule text. Early versions of the draft proposed rule included two
alternative regulatory frameworks. One framework (called ``Framework
A'') offered a licensing approach centered largely on risk analysis and
the other framework (called ``Framework B'') largely replicated the
existing licensing approach in parts 50 and 52 but modified it to be
technology neutral. In its SRM to SECY-23-0021, ``Proposed Rule: Risk-
Informed, Technology-Inclusive Regulatory Framework for Advanced
Reactors (RIN 3150-AK31),'' the Commission disapproved the inclusion of
Framework B in this proposed rule and directed the staff to provide
them within one year an options paper for possible future use of the
Framework B methodology.

B. Stakeholder Views on Part 53 Preliminary Proposed Rule Language

In SRM-SECY-20-0032, the Commission directed the NRC staff to
prepare and release preliminary proposed rule language, followed by
public outreach and dialogue, and then further revise the language
until the NRC staff had established the rudiments of its proposed rule
for Commission consideration. To implement the Commission's direction,
the NRC staff undertook an unprecedented program of stakeholder
engagement, recognizing the importance of this rulemaking to the
advanced reactor community and

[[Page 86921]]

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

III. Discussion

A. Objective and Applicability

The NRC is proposing to add a new, alternative part to its
regulations that would set out a risk-informed, technology-inclusive
framework for the licensing and regulation of commercial nuclear
plants. This new approach would achieve the following: (1) continue to
provide reasonable assurance of adequate protection of public health
and safety and the common defense and security; (2) promote regulatory
stability, predictability, and clarity; (3) reduce requests for
exemptions from the current requirements in parts 50 and 52; (4)
establish new requirements to address non-LWR technologies; (5)
recognize technological advancements in reactor design; and (6) credit
the possible response of some designs of commercial nuclear plants to
postulated accidents, including slower transient response times and
relatively small and slow release of fission products. This proposed
rule would add 10 CFR part 53; subpart M, ``Fitness for Duty Programs
for Facilities Licensed Under 10 CFR Part 53,'' to Part 26; 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 make conforming changes
throughout 10 CFR chapter I, ``Nuclear Regulatory Commission.''

B. Need for Changes to the Existing Regulatory Framework

The NRC has long recognized that the licensing and regulation of a
variety of nuclear reactor technologies would present challenges
because the existing regulatory framework has evolved primarily to
address the LWR designs that compose the current operating fleet
(widely referred to as Generation II reactors). The NRC has had many
interactions with designers of various reactor technologies under
development, sometimes collectively referred to as advanced reactors
(widely referred to as Generation III/III+ (i.e., evolutionary light-
water) and Generation IV (i.e., non-light-water) reactors). The
interactions have informed the development of policies and guidance to
support the potential licensing of new and different types of reactor
facilities, some of which may not utilize LWR designs. The NRC issued
its Advanced Reactor Policy Statement to provide all interested
parties, including the public, with the Commission's views concerning
the desired characteristics of advanced reactor designs. The NRC
further described its 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 is evaluating
its plans for implementing the ADVANCE Act, including how its
regulations, as well as the proposed part 53 or future revisions to it,
could be used to address provisions in the ADVANCE Act. The ADVANCE Act
contains provisions on a variety of nuclear-related topics, such as
micro reactors, nuclear reactor license application reviews, and
nuclear fuel. In Section VI, ``Specific Requests for Comments,'' the
NRC is requesting public input on how part 53 could be revised to
better enable its potential use to implement the ADVANCE Act.
The requirements in part 53 would support a wide variety of
potential commercial nuclear reactor technologies. As noted in this
discussion, the current regulatory framework in parts 50 and 52 evolved
in the context of the current operating reactor fleet dominated by LWRs
and as a result includes provisions specific to LWR technologies. While
the NRC can license other reactor technologies under the current
framework by using existing regulatory flexibilities and the exemption
process, there is significant interest in developing a regulatory
framework that is flexible enough to accommodate multiple technologies
and robust enough to ensure a level of safety equivalent to parts 50
and 52, consistent with the Commission's Advanced Reactor Policy
Statement. The Commission reiterated its safety

[[Page 86922]]

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. In
contrast, part 53 is being developed to accommodate technologies that,
in some cases, lack significant operating experience. To address these
challenges, the NRC drew on well-developed approaches to licensing to
produce a technology-neutral and robust regulatory framework. The
proposed regulatory framework would use PRAs to assess risks, help
establish technical requirements, and manage operations. The framework
builds on the LMP, which is a technology-inclusive approach to
licensing that leverages insights from a detailed PRA to provide
applicants with significant design and operation flexibilities.

C. 10 CFR Part 53: Framework

This proposed rule consists of several major components, including
a new part 53, to be added to 10 CFR chapter I, revisions for part 26,
part 50, and part 73, and conforming changes throughout 10 CFR chapter
I.
Part 53 is comprised of subparts A through M. These provisions are
organized to provide high-level performance criteria and to specify
requirements to demonstrate compliance with those performance criteria
throughout major stages of the life cycle of commercial nuclear plants.
This organization reflects a 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 would complement the design and analysis
requirements in subpart C. The performance-based approach proposed in
part 53 also includes regulatory requirements that would allow
applicants to use a flexible and graded approach to the performance of
safety functions based on the role of a particular SSC, human action,
or program in limiting the overall risks to the public below accepted
standards through balanced measures to prevent and mitigate possible
events.
Proposed subpart M of part 26 would be new and would be largely
consistent with the objective-based fitness for duty (FFD) requirements
in current subpart K, ``FFD Programs for Construction,'' of part 26
supplemented by select requirements from subparts A through I, N, and O
of part 26. These requirements are designed to ensure program
effectiveness, maintain protections afforded to individuals subject to
the FFD program, and align with FFD program implementation by parts 50
and 52 licensees. The proposed requirements are not entirely equivalent
because current subpart K of part 26 only applies during construction
of the commercial nuclear plant, whereas proposed subpart M of part 26
would apply during construction, operation, and decommissioning.
Furthermore, proposed subpart M of part 26 would allow the use of a
variety of biological specimens for drug testing as well as innovative
technologies for drug and alcohol screening and testing that are not
described or allowed by the requirements in subparts A through K, N,
and O of part 26, except under limited conditions.
Proposed revisions to part 73 would establish a new technology-
inclusive consequence-based approach for a range of security areas,
including physical security, cybersecurity, and access authorization
(AA) for commercial nuclear reactors. The NRC used operating experience
to include additional regulatory flexibility for a part 53 licensee's
implementation of security requirements.
In addition, this proposed rule would make conforming changes
throughout 10 CFR chapter I, by adding ``and part 53'' where
appropriate to account for the addition of the proposed part 53.

IV. Part 53: Framework

Subpart A--General Provisions

Subpart A would provide the general provisions applicable to all
applicants and licensees that would be established in part 53 for the
issuance, amendment, and termination of licenses, permits,
certifications, and approvals for commercial nuclear plants licensed
under Section 103 of the Atomic Energy Act of 1954, as amended (the
Act) and title II of the Energy Reorganization Act of 1974 (88 Stat.
1242). Subpart A would include purpose, scope, definitions, written
communications, employee protections, completeness and accuracy of
information, exemptions, standards for review, jurisdictional limits,
consideration of attacks and destructive acts by enemies of the United
States, and information collection requirements.
The requirements in subpart A would be largely equivalent to the
general requirements in part 50 that are applicable to all part 50
applicants and licensees (specifically, Sec. Sec. 50.1 through 50.13)
but would reference the corresponding regulations in part 53 in place
of references to part 50.
A. Discussion of Definitions in Proposed Part 53
This proposed rule would include a definition section in Sec.
53.020. The definitions of most terms in Sec. 53.020 would be
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 proposed 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

[[Page 86923]]

nuclear fission reactor or fusion machine, including a prototype plant
(as defined in sections 50.2 and 52.1 of title 10, Code of Federal
Regulations (as in effect on the date of enactment of this Act)), with
significant improvements compared to commercial nuclear reactors under
construction as of the date of enactment of this Act, including
improvements such as--(A) additional inherent safety features; (B)
significantly lower levelized cost of electricity; (C) lower waste
yields; (D) greater fuel utilization; (E) enhanced reliability; (F)
increased proliferation resistance; (G) increased thermal efficiency;
or (H) ability to integrate into electric and nonelectric
applications.''
Based on public discussions on the use of the term, the NRC
determined that the NEIMA definition, although broad, did not define
``significant improvements'' with enough specificity to implement in
NRC regulations. Additionally, a number of stakeholders suggested that
the descriptor, ``advanced,'' implied enhanced safety, while the NEIMA
definition includes ``significant improvements'' in areas other than
safety enhancements. In response to this feedback, and to be technology
inclusive, the NRC determined that the broader term ``commercial
nuclear plant'' would be preferable. The NEIMA definition of advanced
nuclear reactor also includes fusion technologies. Fusion energy
systems have not been included in the scope of part 53 but are the
subject of a separate rulemaking activity, ``Regulatory Framework for
Fusion Systems.'' See NRC docket ID NRC-2023-0017 on the Federal
rulemaking website <a href="http://www.regulations.gov">http://www.regulations.gov</a>.
The NRC proposes to allow use of part 53 by any ``commercial
nuclear plant.'' The use of the term ``plant'' versus ``reactor,'' as
used in existing regulations (i.e., 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'' was removed from the
definition 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 but is also revised to refer to a ``Commercial nuclear plant'' as
defined in Sec. 53.020.
The NRC proposes to include a definition of ``Consensus code or
standard'' in part 53 that is based on the use of these terms in the
National Technology Transfer and Advancement Act of 1995 (NTTAA) (Pub.
L. 104-113) and the Office of Management and Budget (OMB) Circular No.
A-119, ``Federal Participation in the Development and Use of Voluntary
Consensus Standards and in Conformity Assessment Activities.'' As
required by NTTAA, the NRC undertakes the following activities: (i)
consults with voluntary consensus standards bodies; (ii) participates
with voluntary consensus bodies in the development of consensus
standards; and (iii) uses consensus standards as a means to carry out
the NRC's policy objectives. In part 53, the NRC is not proposing to
incorporate by reference specific codes and standards as is done under
the existing regulations in Sec. 50.55a, ``Codes and standards,''
because some codes and standards are LWR-specific. Part 53 would
require that design features must be designed using generally accepted
consensus codes and standards but would not incorporate the specific
code or standard into the NRC's regulations. During public meetings,
significant discussions with stakeholders indicated that future reactor
designers were interested in the use of international consensus
standards that have not yet been endorsed by the NRC. The definition
proposed in part 53 would allow for the use of international codes and
standards not previously used in NRC licensing but recognizes that the
use of any consensus code or standard would ultimately need to be found
acceptable by the NRC, either through generic efforts to endorse a code
or standard or on an application-specific basis during an individual
licensing review.
The proposed definition of ``Construction'' is slightly different
than the definition in Sec. 50.10--it would cover the same concept but
be applied to a slightly different scope of activities based on how
SSCs are classified under part 53. In part 53, the definition of
``Construction'' is based on the definition in Sec. 50.10 but modified
to apply to safety-related (SR) and non-safety-related but safety-
significant (NSRSS) SSCs identified by the design and analysis
requirements in subparts B and C to ensure the safety criteria are met.
Section 53.020 would also add definitions for terms related to
event selection (LBEs, design-basis accidents (DBAs), anticipated event
sequences, unlikely event sequences, and very unlikely event
sequences); equipment classifications (SR, NSRSS, and non-safety-
significant SSCs); performance metrics (e.g., safety criteria and
functional design criteria); and special treatment.
The regulation would define ``Safety criteria'' in terms of the
plant-level performance-based metrics that would be provided in
Sec. Sec. 53.210 and 53.220. The term ``Functional design criteria''
would be defined as metrics for the performance of specific SSCs that
are determined from the role of the SSC in meeting the safety criteria.
These are new terms that have not previously been defined or used in
NRC regulation.
The term ``Safety-related SSCs'' would refer to those SSCs needed
to meet the safety criteria in Sec. 53.210. The term ``Non-safety-
related but safety-significant SSCs'' would mean 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
``Special treatment'' would be defined as requirements, such as quality
assurance and programmatic controls, identified for each design feature
to ensure that the safety criteria are satisfied and the safety
functions are fulfilled. These requirements would also ensure that SR
and NSRSS SSCs will provide defense in depth, or perform risk-
significant functions, under service conditions and with SSC
reliabilities that are consistent with the analysis required in
proposed subpart C. Structures, systems, and components designated as
SR would also contribute to defense in depth and risk-significant
functions and may warrant special treatments beyond those defined for
the SR functions needed for compliance with Sec. 53.210. The term
``Non-safety-significant SSCs'' would mean those SSCs that are not SR
or NSRSS.
The terms ``Design-basis accidents,'' ``Anticipated event
sequences,'' ``Unlikely event sequences,'' and ``Very unlikely event
sequences'' would be defined to be different types of ``Licensing-basis
events'' and would also be largely equivalent to the LMP's definitions
of DBAs, anticipated operational occurrences (AOOs), design-basis
events (DBEs), and beyond-design-basis events, respectively. The term

[[Page 86924]]

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

Subpart B--Technology-Inclusive Safety Requirements

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

[[Page 86925]]

minimum. The requirements throughout part 53 that support demonstrating
compliance with Sec. 53.220 would be similar to current regulations
that both contribute to assuring adequate protection of public health
and safety and are desirable to promote the common defense and security
or to protect health or to minimize danger to life or property under
Section 161 of the Act.
Consistent with historical practice, Sections 182 and 161 of the
Act are cited as authorizing legislation within this proposed rule.
However, specific language from the Act would not be incorporated into
the safety objectives or safety criteria in part 53. This is because,
again consistent with historical practice, the NRC would not be
defining ``adequate protection'' through the individual safety
requirements in part 53. Rather, part 53 would enable the NRC to make
its required findings under the Act by providing sufficient performance
standards, safety criteria, and related requirements on how applicants
must demonstrate compliance with subpart B and other subparts.
Section 53.210 would provide safety criteria for DBAs that would be
required to be identified under Sec. 53.240 and analyzed under Sec.
53.450(f) in subpart C of part 53. Subsequent sections in part 53 would
require that the SSCs relied upon to demonstrate compliance with the
criteria in Sec. 53.210 be classified as SR. The use of SR SSCs and
the 25 rem reference values for potential radiological consequences
would align with traditional deterministic approaches for LWRs from
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) would be included in Sec. 53.210 to explain
that the use of the 25 rem value would not be intended to imply that
this number constitutes an acceptable limit for an emergency dose to
the public under accident conditions. Rather, this dose value has been
set forth in this proposed section as a reference value that would be
used in the evaluation of plant design features with respect to DBAs to
verify that the proposed designs would provide assurance of low risk of
public exposure to radiation in the event of an accident. The inclusion
of the safety criteria for DBAs in subpart B would provide a logical
structure supporting the identification and treatment of SR SSCs and
establishing the corresponding functional design criteria for those
SSCs.
Section 53.220 would provide safety criteria for LBEs other than
DBAs that would be 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 would provide that a defined
success path exists for DBAs, the safety criteria for LBEs other than
DBAs would establish the connections between SSC design, human actions,
and programmatic controls and a broader set of potential internal and
external hazards. These safety criteria would also address defense-in-
depth matters such as a balanced consideration of prevention and
mitigation.
The safety criterion in Sec. 53.220(b) would include a requirement
to use a comprehensive risk metric or set of metrics and associated
risk performance objectives against which calculated values of the risk
metrics are compared. The comprehensive risk metrics or set of metrics
and associated risk performance objectives would support a performance-
based approach to developing an appropriate combination of design
features and programmatic controls to prevent or mitigate LBEs other
than DBAs. The applicant must propose the comprehensive risk metric or
set of metrics and associated risk performance objectives, and the
comprehensive risk metric or set of metrics and associated risk
performance objectives must provide an appropriate level of safety.
Comprehensive risk metrics should consist of a proposed plant risk
metric or set of proposed risk metrics that approximate the total,
overall risk from the facility and that address the range of possible
plant configurations and associated internal and external hazards to
the extent practicable. The associated risk performance objectives are
preestablished, indicative values of the comprehensive risk metrics
that are used as part of risk-informed decision-making. The methodology
for developing and using proposed comprehensive risk metrics and
associated risk performance objectives is defined by the proposed
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 would not constitute a real-time
requirement that must be continuously demonstrated by the licensee. The
use of a comprehensive risk metric or set of risk metrics and risk
performance objectives that reflect an average risk to establish
performance goals for SR and NSRSS SSCs is consistent with current
practices that use other risk assessment techniques to address short-
term plant configurations during plant maintenance activities.
It is worth noting that the evaluation of plant risks, as
represented by a comparison of analysis results to acceptable risk
performance objectives for comprehensive risk metrics, would be one of
several performance standards used in subpart B. The proposed use of
multiple performance standards, including deterministic criteria and
defense-in-depth measures, reflects an integrated decision-making
process similar to that described in RG 1.174, ``An Approach for Using
Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-
Specific Changes to the Licensing Basis,'' Revision 3. The NRC's
approval of using a comprehensive risk metric or set of metrics with
associated risk performance objectives is not, by itself, an indicator
of adequate protection. Rather, the comparison of comprehensive risk
metrics to associated risk performance objectives that are acceptable
to the NRC is part of a suite of regulatory requirements that, when
considered holistically, form the basis for the NRC's decision-making.
This is analogous to the approach used for plants licensed under part
50 and part 52, where no single regulatory requirement governs whether
a plant is ``safe enough.''
The RG 1.233, ``Guidance for a Technology-Inclusive, Risk-Informed,
and Performance-Based Methodology to Inform the Licensing Basis and
Content of Applications for Licenses, Certifications, and Approvals for
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

[[Page 86926]]

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 proposed requirements in subparts B and C, would bring the
approach endorsed in RG 1.233 for parts 50 and 52 into part 53.
Additionally, the use of comprehensive risk metrics and associated risk
performance objectives would provide a logical performance objective to
support the risk management approaches in the various subparts
comprising proposed part 53.
The Commission stated in the introduction of the Safety Goals
Policy Statement that improvements to 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) could be expressed in terms of a
biologically average individual in terms of age and other risk factors.
Although some comprehensive risk objectives such as the IEFR and ILCFR
are defined in terms of fatality risks, the Commission continues to
make clear that no death attributable to nuclear power plant operation
will ever be ``acceptable'' in the sense that the Commission would
regard it as a routine or permissible event. Comprehensive risk metrics
and associated risk performance objectives as used in this proposed
rule would establish acceptable risks, not acceptable deaths.
Applicants under the proposed part 53 may choose to develop and
seek NRC approval of comprehensive risk metrics or sets of risk metrics
and associated risk performance objectives beyond those discussed
above, including the use of surrogate measures for use in specific
analyses to satisfy the proposed 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 would, as appropriate, review novel approaches for
comprehensive metrics and associated risk performance goals proposed by
applicants, industry organizations, or standard development
organizations and would engage stakeholders during the development of
the related regulatory guidance or specific licensing actions.
Section 53.230 would require safety functions needed to ensure that
the safety criteria under 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
would specify that limiting the release of radioactive materials from
the facility is the primary safety function, and therefore, limiting
potential offsite consequences (i.e., dose to a hypothetical
individual) would be used as the primary performance metric throughout
part 53. The additional or subsidiary safety functions needed to limit
the release of radionuclides may include, without limitation,
controlling processes related to reactivity, heat generation, heat
removal, and chemical interactions. This proposed rule provides
flexibility to applicants and licensees in identifying, implementing,
and maintaining the safety functions supporting retention of
radionuclides for commercial nuclear plants of varying sizes and
technologies.
Proposed Sec. 53.240 would require applicants to identify and
address LBEs. LBEs are unplanned events, resulting from both internal
and external hazards, that are used in the design and analyses required
under part 53 for licensing commercial nuclear plants. This ensures
estimates of offsite consequences from analyses performed under
proposed Sec. 53.450 are below the safety criteria identified under
proposed Sec. Sec. 53.210 and 53.220 and that SSCs, personnel, and
programs address the safety functions from proposed Sec. 53.230.
Including a high-level performance requirement related to the
identification and analysis of LBEs in subpart B would reflect the
historical and continuing importance of evaluating unplanned events as
part of the licensing of commercial nuclear plants. Proposed Sec.
53.240 would require identification and analysis of LBEs under Sec.
53.450, which would require a PRA. Examples of acceptable methods of
using PRAs to identify and assess LBEs would be the methodology in RG
1.233, as discussed in Draft Regulatory Guide (DG)-1413, ``Technology-
Inclusive Identification of Licensing Events for Commercial Nuclear
Plants.''
Section 53.250 would establish defense-in-depth requirements based
on the longstanding philosophy of providing defense in depth to address
uncertainties about the design, operation, and performance of
commercial nuclear plants. For example, parts 50 and 52 address defense
in depth through layered prescriptive technical requirements (e.g.,
fuel performance, cladding integrity, reactor coolant system integrity,
containment performance) for LWRs. In contrast, the flexibility
afforded to applicants in how they propose to demonstrate compliance
with the high-level safety criteria within part 53 would necessitate
this specific requirement to ensure defense in depth is provided. The
requirements in this section would state that no single engineered
design feature, human action, or programmatic control, no matter how
robust, should be exclusively relied upon to address LBEs other than
DBAs. The phrase ``engineered design feature'' would not preclude the
possible crediting of inherent characteristics within the design and
analysis for commercial nuclear reactors. While defense in depth would
only be assessed for LBEs other than DBAs, the need to ensure dedicated
success paths for DBAs would contribute to the overall defense in depth
for each commercial nuclear plant under part 53.
Section 53.260 would govern normal operations and would establish a
level of safety based on current requirements in 10 CFR part 20,
``Standards for Protection Against Radiation,'' which limits doses to
members of the public and dose rates in unrestricted areas.
Section 53.270 would provide for the protection of plant workers
and would establish a level of safety based on current requirements in
10 CFR part 20 which limits occupational dose.

Subpart C--Design and Analysis Requirements

This subpart would provide requirements for the design of
commercial nuclear plants and the supporting analyses, including the
analyses of LBEs, to demonstrate that the performance standards in
proposed

[[Page 86927]]

subpart B can be satisfied. The sections within subpart C would reflect
the overall hierarchy throughout part 53, which would cover: (1) plant-
level safety criteria (Sec. Sec. 53.210, 53.220, and 53.470); (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 on to demonstrate the safety
criteria (Sec. Sec. 53.210, 53.220, and 53.470) are met. Subpart C
would also contribute to the logic and structure of part 53 by
distinguishing between SR SSCs and NSRSS SSCs and 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. Non-safety-
related but safety-significant SSCs, human actions, and licensee-
controlled programs that address LBEs other than DBAs generally
contribute to the appropriate measures considering potential risks to
public health and safety.
Section 53.400 would establish a requirement that design features
be provided for each commercial nuclear plant to satisfy the safety
criteria and fulfill safety functions from proposed subpart B during
LBEs. Other sections in subpart C would, in turn, further address the
necessary capabilities and reliabilities for SSCs by establishing
functional design criteria, fulfilling design requirements, performing
analyses of LBEs, performing other supporting analyses, and
categorizing SSCs based on their roles in preventing or mitigating
LBEs.
Section 53.410 would require that functional design criteria be
defined for design features relied upon to demonstrate that the
consequences from DBAs would be below the criteria in 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 would establish appropriate controls on these design
features (e.g., safety classification, protection from external
hazards, quality assurance, and TS) to ensure the functional design
criteria are satisfied. The performance requirements for the SSCs
needed to address DBAs and the corresponding human actions and
programmatic controls would contribute to ensuring that a commercial
nuclear plant licensed under part 53 would meet the safety criteria in
Sec. 53.210.
Section 53.415 would require that SR SSCs be protected against or
designed to withstand the effects of natural phenomena (e.g.,
earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches) and
constructed hazards (e.g., from dams, transportation routes, and
military or industrial facilities). Specifically, Sec. 53.415 would
require 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 would refer to
such things as the magnitude and recurrence rate of an earthquake and
the resultant ground motions, the height of a flood, the force of
hurricane winds, or the concentrations of chemicals resulting from a
release from a nearby facility. These requirements would support either
traditional deterministic approaches for determining and protecting
against external hazards or probabilistic approaches that are being
developed for seismic and some other external hazards.
Section 53.420 would require that functional design criteria be
defined for design features that play a significant role in
demonstrating that the safety criteria for LBEs other than DBAs are
satisfied. The analyses required for this demonstration would be
described in proposed Sec. 53.450(e), which would require that those
events be identified and assessed using a PRA methodology in
combination with other generally accepted approaches for systematically
evaluating engineered systems. The SSCs determined to be safety
significant (i.e., either SR or NSRSS) would have associated special
treatment requirements as specified in Sec. 53.460. Special treatment
would be defined in subpart A of part 53 and generally refers to
measures (e.g., quality assurance, testing, monitoring) taken beyond
the procurement and installation of commercial grade products to
provide confidence that the SSC will comply with the applicable
functional design criteria. The inclusion of a systematic approach to
identifying the functional design criteria for SSCs and tailoring the
special treatments to specific LBEs and safety functions is an
important contributor to satisfy the proposed safety criteria in
subpart B. Therefore, designers and licensees for commercial nuclear
plants would be provided flexibility on how LBEs other than DBAs are
either prevented or mitigated and how the calculated comprehensive
plant risks satisfy the safety criterion established under Sec.
53.220(b).
Section 53.425 would establish requirements for design features and
related functional design criteria limiting doses to members of the
public during normal operations to satisfy the criteria in part 20.
Section 53.430 would provide similar requirements for design features
and related functional design criteria for protection of plant workers
to meet the safety criteria in part 20. Similar to existing
regulations, the NRC considers that licensees would generally comply
with the requirements of part 20 to keep doses as low as reasonably
achievable by meeting a design objective of keeping doses to the public
from routine plant effluents less than 10 millirem per year. This goal
is similar to that provided by appendix I to part 50 and would assist
designers, applicants, and licensees in performing the evaluations of
possible reductions in public dose from routine effluents when
considering costs and other factors. As emphasized in existing
regulations in part 50, the design objective of keeping doses to the
public from routine plant effluents less than 10 millirem per year
should not be construed as a radiation protection standard. The NRC
anticipates that future guidance will continue to reflect this
performance goal.
The proposed requirements in 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 would address 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 would be met through
design practices, consideration of testing and operating experience,
and various assessments of LBEs and other potential challenges to
commercial nuclear plants. Discussions of some of the key design
requirements included in this section follow.
<bullet> Sec. 53.440(a): An essential element to ensuring a
proposed design can comply with the performance criteria in proposed
part 53 would be that the abilities of design features to fulfill their
safety functions are demonstrated by a combination of analyses, test
programs, prototype testing, and operating experience. This requirement
closely aligns with the language in Sec. 50.43(e)

[[Page 86928]]

and is proposed in part 53 as the same foundational requirement. In
addition, the proposed Sec. 53.440(a) would require the design
processes for SSCs under this section to include administrative
procedures for evaluating operating, design, and construction
experience for considering applicable important industry experiences in
the design of those SSCs. This proposed requirement corresponds to the
existing requirement under Sec. 50.34(f)(3)(i) that was developed in
response to the 1979 accident at Three Mile Island Nuclear Generating
Station.
<bullet> Sec. 53.440(b): The design and licensing of commercial
nuclear plants should use generally accepted consensus codes and
standards. Such codes and standards ensure sufficient testing and
qualification of materials and equipment and provide defined processes,
specifications, and acceptance criteria for use by designers and
suppliers. The NRC would indicate acceptance of consensus codes and
standards used in the design and licensing of a specific commercial
nuclear plant either through the NRC's generic endorsement of a code or
standard (i.e., through regulatory guidance), including any limitations
or conditions, that can be referenced within an application, or through
the review of a referenced code or standard as part of the review of a
specific application.
<bullet> Sec. 53.440(c): The design requirements in subpart C
would require the materials used for SR and NSRSS SSCs to be qualified
for their service conditions over the design life of the SSC.
<bullet> Sec. 53.440(d): The requirements in Sec. 53.440 would
include the need to consider possible degradation mechanisms for
materials and equipment to inform both the design process and the
development of integrity assessment programs to be executed during
plant operations in accordance with subpart F of part 53. The inclusion
of requirements related to designing and monitoring for possible
degradation mechanisms reflects important lessons learned from the
history of LWRs as well as operating experience with structures and
systems in countless other engineering endeavors.
<bullet> Sec. 53.440(e) and (f): The design requirements in
subpart C would state specific design requirements similar to existing
requirements in parts 50, 52, and 73 for protections against fires and
explosions and consideration of safety and security together in the
design process.
<bullet> Sec. 53.440(g) and (h): Specific design requirements are
proposed to ensure that commercial nuclear reactors under part 53 have
the capability to achieve and maintain subcriticality and long-term
cooling. The requirements would be included to address the potential
that some reactor designs may be able to achieve a stable end state for
the purpose of event analyses but might need further actions to
completely shut down and service the facility.
<bullet> Sec. 53.440(i): The design, analysis, and development of
programmatic controls under part 53 would consider the number of
reactor units and other significant inventories of radioactive
materials contributing to the risks to public health and safety. This
would reflect the definition of ``Commercial nuclear plant'' in subpart
A and reinforce that the evaluation of LBEs is performed on a plant-
wide basis. This aspect of part 53 would be different from parts 50 and
52, which generally define safety requirements on the assumption of
events involving only individual reactor units.
<bullet> Sec. 53.440(j): A design requirement is proposed to
provide a technology-inclusive requirement that would be equivalent to
the requirements in Sec. 50.150 to address the possible impact of a
large commercial aircraft.
<bullet> Sec. 53.440(k): The inclusion of a specific proposed
requirement to address the risks to public health from potential
chemical hazards of licensed material is appropriate given the
diversity of reactor technologies and designs that might be licensed
under part 53. The requirement in part 53 would be similar to the
existing requirements in 10 CFR part 70, ``Domestic Licensing of
Special Nuclear Material,'' that address both potential radiological
and chemical hazards for licensed materials at fuel cycle facilities.
<bullet> Sec. 53.440(l): Provisions are proposed to require that
measures be taken during the design of commercial nuclear plants to
minimize contamination of the facility and the environment, facilitate
eventual decommissioning, and minimize the generation of radioactive
waste in accordance with Sec. 20.1406.
<bullet> Sec. 53.440(m): A design requirement is proposed to
provide 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 handling and storage that
would prevent inadvertent criticality events.
<bullet> Sec. 53.440(n): The design would need 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.
Section 53.450 would establish analysis requirements and would
center upon the use of a PRA in combination with other generally
accepted approaches for systematically evaluating engineered systems.
The reliance on PRAs as a key component in the proposed analysis
requirements for part 53 would reflect the decades of improvements in
PRA methodologies and the increasing use of PRA techniques in the
design, licensing, and oversight of both operating and future nuclear
reactors. Part of the Commission's PRA Policy Statement is that the use
of PRA technology should be increased in all regulatory matters to the
extent supported by the state of the art in PRA methods and data and in
a manner that complements the NRC's deterministic approach and supports
the NRC's traditional defense-in-depth philosophy. The need to
supplement PRA insights with other engineering approaches and judgments
reflects the NRC's longstanding policy described in the SRM to SECY-98-
144, ``Staff Requirements--SECY-98-144--White Paper on Risk-Informed
and Performance-Based Regulations,'' dated February 24, 1999, for
regulatory decision-making to be risk-informed but not solely based on
numerical results of a risk assessment (i.e., not a risk-based
approach). Part 53 would maintain a role for NRC's traditional
deterministic approaches (particularly for DBAs) and defense-in-depth
philosophy by including specific requirements utilizing these
regulatory tools in subparts B and C.
PRA would be used in combination with other techniques in part 53
to identify and categorize LBEs, classify SSCs, and evaluate defense in
depth. This increased role for the PRA necessitates that it would be
developed, performed, and maintained in accordance with NRC-approved
standards and practices (see 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 would need to be qualified for the
range of conditions being simulated across a wide range of unplanned
events. These analyses would need to use realistic approaches and
address uncertainties associated with states of knowledge, modeling,
and performance of SSCs.
While industry consensus PRA standards and peer review processes
endorsed in RGs 1.200 and 1.247 remain

[[Page 86929]]

acceptable for developing a PRA, they are not regulatory requirements
and an application under part 53 need not follow every aspect of the
applicable consensus PRA standard. Existing processes for defining the
scope and capability of a PRA supporting an application offer
flexibility in determining the degree to which the PRA needs to be
developed and may be informed by other factors such as design
complexity and the needed degree of realism and level of detail,
consistent with the use of the PRA and substance of the application.
Such processes are currently available for appropriately defining the
scope of the PRA and determining applicability of supporting
requirements in consensus PRA standards needed to satisfy the proposed
regulatory requirements for the specific uses of analyses under Sec.
53.450(b). 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.
The proposed Sec. 53.450(c) would require periodic maintenance and
upgrading of the PRA to maintain an alignment between the supporting
analyses and the design and performance of plant equipment, programs
and procedures, and other factors associated with meeting the safety
criteria of the proposed Sec. 53.220 and the evaluation criteria of
proposed Sec. 53.450(e)(2). The periodic maintenance of the PRA would
also be a means to consider new or revised information related to
external hazards, industry operating experience, performance issues
with or degradation of SSCs, and other contributors to the frequency
and potential consequences of various event sequences. The periodic
assessments performed by licensees to support the maintenance of the
PRA and other requirements in the proposed part 53 would be
complemented by NRC inspections and programs to assess new or revised
information related to topics such as natural hazards, operating
experience, and potential generic safety issues.
The categories of LBEs used in part 53 would include anticipated
event sequences, unlikely event sequences, and very unlikely event
sequences. The unlikely event sequences would include those events with
estimated frequencies well below the frequency of events expected to
occur during the lifetime of a commercial nuclear plant. An important
aspect of the analysis requirements is that, under proposed Sec.
53.450(e), the analyses of LBEs other than DBAs would not only be used
to show the performance criteria of Sec. 53.220 are satisfied but to
also show that evaluation criteria defined for each LBE or category of
LBEs would also be satisfied. Such evaluation criteria for specific
LBEs or categories of LBEs would be defined in terms of limits on the
release of radionuclides or maintaining the integrity of one or more
barriers used to limit the release of radionuclides and reflect a
graded approach of allowing lesser potential consequences from more
frequent events. An example of such evaluation criteria for a range of
LBEs that could likely be expanded for part 53 is provided in RG 1.233.
Another proposed requirement for the proposed Sec. 53.450(e) analyses
is that the methodology would need to include a means to identify event
sequences deemed risk-significant such that those event sequences can
be given special attention within other sections of part 53.
Part 53 would maintain an important role for a deterministic
analysis of DBAs in the performance criteria of Sec. 53.210 and the
related analytical requirements in Sec. 53.450(f). The analysis of
DBAs would be required to address event sequences drawn from those with
estimated frequencies below the expected lifetime of a generation of
reactors (e.g., event sequences with frequencies as low as one in ten
thousand years). As proposed in this section, DBAs would need to be
analyzed using deterministic methods and ensure a safe, stable end
state with reliance upon only SR SSCs and human actions, if needed, to
be performed by operators licensed under the provisions of Sec. Sec.
53.760 through 53.795.
While the DBAs analyzed under part 53 would be similar to the
traditional DBAs analyzed under parts 50 and 52, there are important
distinctions between the overall role of DBA analyses in part 50 and
proposed part 53. In part 53, the role of the DBA analysis would be
more narrowly focused on selecting SR SSCs and determining functional
design criteria for those SSCs to ensure the commercial nuclear plant
meets the safety criteria in Sec. 53.210. The overall control of risks
posed by commercial nuclear plants under part 53 would be provided by
the analyses of and measures taken for both DBAs and other LBEs,
including very unlikely event sequences. This would contrast with the
traditional deterministic approach in part 50 wherein the analyses of
DBEs such as DBAs were used to provide bounding assessments,
incorporate standard design rules such as assumptions related to single
failures, and to define conservative performance requirements for SR
SSCs. Limitations related to the traditional deterministic approach
were addressed in part 50 through case-by-case assessments and specific
actions for beyond-design-basis events such as anticipated transients
without scram and station blackout.
Section 53.450 would also include provisions to ensure that
analyses are performed to support the design requirements of Sec.
53.440(e) on fire protection, Sec. 53.440(j) on aircraft impact
assessments, 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 proposed
analysis requirements related to fire protection would support either a
traditional, deterministic approach or a more risk-informed approach
where the risks from fires are addressed within the identification and
analyses of LBEs.
Section 53.460 would establish criteria for the safety
classification of SSCs and determination of appropriate special
treatments. As noted in subpart A, the term ``Special treatments''
would be defined to mean those items, such as measures taken to satisfy
functional design criteria, quality assurance, and programmatic
controls, which provide assurance that certain SSCs will provide
defense in depth or perform risk-significant functions. These
requirements would also provide confidence that the SSCs will perform
under the service conditions and with the reliability credited in the
analysis performed in accordance with Sec. 53.450 to satisfy the
safety criteria in Sec. Sec. 53.210 and 53.220. The terminology used
in part 53 would include the following categories for SSC
classification: (1) SR; (2) NSRSS; and (3) non-safety significant.
Requirements for SR SSCs would be defined in other sections of

[[Page 86930]]

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

Subpart D--Siting Requirements

Proposed subpart D in part 53 would state requirements for the
siting of commercial nuclear plants and would serve the role provided
by 10 CFR part 100, ``Reactor Site Criteria,'' for nuclear reactors
licensed under parts 50 and 52. As reflected in proposed Sec. 53.500,
the reason for establishing siting requirements would remain the same
as it has been historically, which is to ensure that licensees and
applicants assess what impact the site environs may have on a
commercial nuclear plant (e.g., external hazards) and, conversely, what
potential adverse health and safety impacts a commercial nuclear plant
may have on nearby populations in view of the site characteristics.
Proposed Sec. 53.510 would require that design-basis external
hazard levels be identified and characterized based on site-specific
assessments of natural and constructed hazards with the potential to
adversely affect plant functions. The site-specific assessments would
be used in the proposed Sec. 53.415, which would require that SR SSCs
be designed to withstand the effects of natural phenomena and
constructed hazards of levels or severities up to design-basis external
hazard levels. The design-basis levels for external hazards relevant to
a site would need to account for uncertainties and variabilities in
data, models, and methods used to characterize those hazards. Existing
approaches could be used to demonstrate compliance with this
requirement. The historical importance of assessing seismic events as
risks to commercial nuclear plants and the associated development of
risk-informed approaches to address seismic events would be reflected
in proposed 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,'' would be an
acceptable way to develop site-specific ground motion response spectra
for SSCs under appendix S to part 50, which corresponds to SSCs that
are categorized as the highest SDC (SDC-5) in ASCE/SEI 43-19.
The evaluation of seismic hazards under subpart D would need to be
sufficient to inform a site-specific design (e.g., a CP or custom COL)
or confirm the use of a standard design for a commercial nuclear plant
under Sec. 53.480 and other sections of subpart C. A risk-informed
approach could use several design-basis ground motions (DBGMs) to
assess SSCs in various SDCs (i.e., one DBGM per SDC). Section 53.510(d)
would state that geologic and seismic siting factors must also include
related hazards such as seismically induced flooding and volcanic
activity that may affect the design and operation of a proposed
commercial nuclear plant for the proposed site.
Section 53.520 would require applicants to identify and assess site
characteristics related to topics which might include meteorology,
geology, hydrology, or other areas in the design and analyses required
under subpart C.
Proposed section 53.530 would set requirements for population-
related considerations and maintain requirements and definitions
similar to those currently in part 100 for an exclusion area, low
population zone,

[[Page 86931]]

and population center distance. The NRC recognizes that some applicants
may propose to essentially collapse the exclusion area and low
population zone to the site boundary. This approach would rest on a
demonstration that the calculated consequences of DBAs remain below the
proposed dose guidelines used in Sec. 53.210, which are the same as
those in the existing regulations in parts 50, 52, and 100. The
proposed definitions in Sec. 53.020 would allow such configurations,
assuming they were justified by the design and analyses from subpart C.
This approach should provide flexibility to justify alternative
exclusion areas and low population zones without foreclosing the option
for an applicant to define more conventional exclusion areas and low
population zones outside of a defined site boundary. The NRC's long-
standing preference for siting reactors in areas of low population
density would be maintained in part 53 by using the current language
from part 100 in proposed 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.'' Site-related requirements in part 20 (restricted area) and
part 73 (protected and owner-controlled areas) would remain applicable
to commercial nuclear plants licensed under part 53.
Proposed section 53.540 would require that site characteristics be
appropriately considered in other activities such as the design and
analysis performed under proposed subpart D and the emergency planning
and security programs under proposed subpart F.

Subpart E--Construction and Manufacturing Requirements

The proposed part 53 language would establish construction and
manufacturing requirements in subpart E. The proposed language for
construction-related activities would largely reflect current
requirements in part 50 without any fundamental changes. Limited
changes would be made in several places, as described in the following
paragraphs, to be technology-neutral and for consistency with the
organization and language of part 53. The proposed language for
requirements for manufacturing activities would largely mirror those
for construction-related activities. However, the proposed
manufacturing requirements have been updated from the current
requirements in subpart F of part 52 to better accommodate the possible
factory fabrication of manufactured reactors. The manufacturing of
specific components outside the scope of an ML would not be addressed
by these proposed subparts.
Section 53.600 would establish the overall construction and
manufacturing requirements for CPs, OLs, COLs, MLs, and limited work
authorizations (LWAs). This section would connect the construction and
manufacturing requirements to the safety criteria, quality assurance
requirements, and other requirements located in other subparts. These
requirements would require that construction and manufacturing
activities be managed and conducted such that when combined with
associated design features and programmatic controls, the constructed
plant would satisfy the relevant requirements in subpart B.
Section 53.605 would establish requirements for the reporting of
defects and instances of noncompliance during construction. This
section would provide equivalent requirements to those in Sec.
50.55(e).
Section 53.610(a) would establish the requirement to have in place
a well-defined command and control structure to manage construction
activities. The requirements would generally reflect current
requirements, with an emphasis on the quality assurance programs for
complying with the requirements in appendix B to part 50. The proposed
Sec. 53.610(a)(6) would require programmatic controls for implementing
special treatment for NSRSS SSCs to align with requirements in other
subparts in part 53. The section would also refer to other NRC
regulations to address matters such as requirements to have a FFD
program, a radiation protection program if radioactive materials are
brought onto the site, and security programs to protect sensitive
information and protect against cyber threats.
Section 53.610(b) would provide requirements governing construction
activities, including the equivalent of the requirement in 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) would require procedures to be in place prior to
beginning construction to ensure that construction-related activities
do not undermine important features such as slope stability and that
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 would be equivalent to
requirements in parts 50 and 52 with appropriate references to other
parts for items such as possession of byproduct material or SNM,
protecting operating units from construction activities for commercial
nuclear plants with multiple reactor units, and having a redress plan
in case LWA activities are terminated.
Section 53.610(c) would address inspection and acceptance
activities by including requirements in part 53 equivalent to specific
quality assurance criteria in appendix B to part 50 and inspections,
tests, analyses, and acceptance criteria (ITAAC) in part 52 for COLs.
Section 53.620(a) would include proposed requirements covering the
activities performed under an ML issued under part 53. Provisions
related to MLs were first adopted by the NRC in 1973 through the
addition of appendix M to part 50. The regulation supported the
manufacture of a nuclear power reactor to be incorporated into a
commercial nuclear plant under a CP and operated under an OL at a
different location from the place of manufacture.\1\ The regulations
and processes for MLs were changed substantially in the part 52
rulemaking in 2007 (72 FR 49352). The most important shift in the ML
concept in that rulemaking was that a final reactor design, which would
be equivalent to that required for a standard DC under part 52 or an OL
under part 50, must be submitted and approved before issuance of an ML.
The rationale for that change was that approval of a final design
ensures early consideration and resolution of technical matters before
there is any substantial commitment of resources associated with the
actual manufacture of the reactor, which greatly enhances regulatory
stability and predictability.
---------------------------------------------------------------------------

\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 proposed part 53 sections in subpart E for manufacturing and in
subpart H for licensing matters would maintain requirements equivalent
to those in part 52 for MLs. The NRC approval of a standard design and
related manufacturing processes, coupled with a stable workforce and
established procedures, has the potential for maintaining and even
improving the quality and consistency of manufacturing, as compared to
the traditional method of constructing

[[Page 86932]]

reactors onsite by a variety of contractors and subcontractors.
Subpart E would include requirements that would apply to portions
of a manufactured reactor in recognition that some activities covered
by an ML may occur at different fabrication facilities. As with the
preceding sections on construction, Sec. 53.620 would establish the
requirements to have in place programs, procedures, and a well-defined
command and control structure to manage manufacturing-related
activities.
Section 53.620(b) in subpart E would propose requirements for
executing the manufacturing activities following receipt of an ML under
part 53. Information about the design and manufacturing processes
should be provided by the applicant. The importance of the ML is
reflected in several of the proposed requirements in Sec. 53.620(b)
that would refer to complying with the ML, including conducting
manufacturing processes within facilities for which the license holder
can control activities. The essential role of post-manufacturing
inspections would also be incorporated into this proposed section by
requiring the holder of the ML to perform inspections and have
acceptance processes for manufactured reactors or portions of a
manufactured reactor.
Section 53.620(c) would provide proposed requirements for the
control of radioactive materials if the holder of an ML plans to
possess and use source, byproduct, or SNM as part of the manufacturing
process. By and large, the proposed subpart E would refer to NRC
regulations in 10 CFR part 30, ``Rules of General Applicability to
Domestic Licensing of Byproduct Material,'' 10 CFR part 40, ``Domestic
Licensing of Source Material,'' and part 70 for the requirements on
controlling radioactive materials. Several specific requirements to
address the potential hazards of radioactive materials are proposed in
areas such as having a fire protection program, an emergency plan,
training programs, and procedures to minimize contamination.
The most significant change proposed for MLs in part 53 as compared
to MLs under part 52 relates to Sec. 53.620(d) in subpart E and the
associated licensing provisions in subpart H. These provisions would
allow and establish requirements for the loading of fuel into a
manufactured reactor at the manufacturing site for subsequent transport
to a commercial nuclear facility that will operate pursuant to a COL.
The first requirement in the proposed Sec. 53.620(d) would establish
limitations on when a license under part 70 would authorize the loading
of fuel into a reactor manufactured under an ML. The proposed
regulation would require the manufactured reactor to include at least
two independent physical mechanisms that will each prevent criticality
should conditions most favorable to critical operation be introduced
(e.g., optimum neutron moderation and reflection). This requirement
would contribute to the NRC's longstanding practice of requiring
defense in depth for preventing accidents in any facility dealing with
SNM, including requirements in Sec. 70.64 for certain part 70
licensees to adhere to the ``double contingency principle.''
The requirements to have in place mechanisms to prevent criticality
could likewise support meeting other provisions in subpart H to part
70, such as those related to having a safety program and integrated
safety assessment. The mechanisms to preclude criticality in the
proposed requirements would reasonably ensure that a manufactured
reactor would not become critical assuming optimum neutron moderation,
and optimum neutron reflection conditions. With the proposed
requirements for mechanisms to prevent criticality and all criticality
safety controls required by 10 CFR part 70 in place, the presence of
fuel in the manufactured reactor would not create a nuclear hazard
different than the hazard from the presence of the same fuel in a
storage location or container licensed under 10 CFR part 70.
Collectively, the proposed measures would reasonably ensure that the
manufactured reactor would not be capable of operations, thereby
obviating the need for a COL under Sec. Sec. 53.1416 and 53.1440 to
authorize fuel loading. Additionally, this approach would focus the ML
application and its review on the design, manufacture, and deployment
of the manufactured reactor.
The activities involving SNM within the manufacturing facility,
including the loading of fuel, would be regulated primarily under the
part 70 license. The reference to the requirements in subpart H of part
70 in section 53.620(d) assures that the activities involving the
receipt, storage, and loading of a variety of possible fuel forms and
enrichments at the manufacturing facility will be analyzed in a
systematic manner and appropriate protection will be provided against
equipment malfunctions, human errors, external hazards, and other
adverse conditions. The regulations in part 51 provide a flexible
approach for environmental review to address the range of regulated
activities under part 70. The flexibility in part 51 will enable the
NRC to determine the appropriate type of environmental review based on
the circumstances associated with the loading of fuel into a specific
manufactured reactor.
The proposed Sec. 53.620(d) cites the requirements in parts 70,
71, and 73 to ensure important features and programs are in place prior
to the receipt of SNM. The features and programs required to be in
place prior to receipt of SNM include (1) radiation monitoring
instrumentation and alarms; (2) measures to detect potential
criticality accidents; (3) appropriate procedures, equipment, and
personnel qualified for the fuel loading; (4) programs for physical
security and cybersecurity; and (5) material control and accounting
(MC&A) programs. Section 53.620(d)(2)(i) proposes requirements to
address security programs for any ML authorizing possession of a
manufactured reactor into which fuel has been loaded at the
manufacturing facility. Currently, for category II SNM, security
measures may be required in addition to requirements included in 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 proposed part 53 regulations will
provide additional openness and transparency for applicants applying
for an ML who seek to load fuel into manufactured reactors at a
manufacturing site.
Currently, 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 proposed physical security program
for fueled manufactured reactors would require a security plan for any
ML authorizing possession of a manufactured reactor into which fuel has
been loaded at the manufacturing facility, regardless of fuel type,
enrichment, and quantity. This is consistent with other controls for
MLs, including reactivity and criticality controls.
The proposed requirements would also require a holder of an ML and
part 70 license to address cybersecurity to ensure a cyberattack would
not adversely impact the functions performed by digital assets used by
the licensee for physical security, radiation monitoring, or
criticality prevention.
The proposed regulations in part 53 covering the activities related
to the storage, movement, and loading of fresh

[[Page 86933]]

fuel into a manufactured reactor in the manufacturing facility would
likewise refer to the applicable regulations in part 70. The proposed
Sec. 53.620(d) would also require the loading or unloading of
unirradiated fuel into or from a manufactured reactor and any changes
to the configuration of reactivity-related systems to be performed by a
certified fuel handler meeting the requirements in subpart F. The NRC
is aware of proposals to introduce reprocessing of existing or future
spent nuclear fuel into the fuel cycle for some potential commercial
nuclear plants. This proposed rule does not address the loading of
spent nuclear fuel or fuel resulting from reprocessing of spent nuclear
fuel into a manufactured reactor.
Section 53.620(e) would limit the transport and delivery of a
manufactured reactor or portions of a manufactured reactor only to a
site for which the Commission has issued a COL authorizing the
construction of a commercial nuclear plant using a manufactured reactor
under the specific ML. This proposed requirement is similar to the
limitations in Sec. 52.153, with the difference being that part 53
would allow the installation of a manufactured reactor at the site of a
COL but would not include provisions for installation at a site under a
CP. The possible combination of a manufactured reactor and the
licensing option of CP and OL seems unlikely and would require the
introduction of ITAAC into the licensing provisions for a CP and OL. An
additional proposed paragraph in Sec. 53.620(e) would provide
requirements for protecting fueled manufactured reactors during
transport to the site of the commercial nuclear plant by referencing
the transportation and security requirements in 10 CFR part 71,
``Packaging and Transportation of Radioactive Material,'' and part 73.
Section 53.620(f) would include proposed requirements for the
acceptance and installation of a manufactured reactor at the site of a
commercial nuclear plant. The proposed requirements would reference the
construction requirements in Sec. 53.610 to govern the integration of
the manufactured reactor into the construction of a commercial nuclear
plant. Other proposed requirements in the section would address
required receipt inspections and verification that interface
requirements between the manufactured reactor and the balance of the
commercial nuclear plant have been met.

Subpart F--Requirements for Operation

Proposed subpart F would provide the requirements for the
operations phase of a commercial nuclear plant to ensure that the
safety criteria in subpart B are satisfied throughout the plant's
lifetime and during all modes of normal operation and unplanned events.
Section 53.700 would provide the overall objectives and general
organization of subpart F, which would be to establish requirements
during operations for: (1) plant SSCs; (2) plant personnel; and (3)
plant programs.
Proposed Sec. 53.710 would provide the requirements for
maintaining capabilities, availability, and reliability of SSCs to
demonstrate compliance with the safety criteria and design requirements
for unplanned events that are described in proposed subparts B and C.
The basic structure of this proposed section would be that controls for
SR SSCs are provided by TS and controls for NSRSS SSCs are required to
be addressed with licensee-controlled documents and procedures.
The general content and control of TS under the proposed part 53
would be similar to the requirements in part 50. The proposed
requirements for TS would include limits on the inventories of
radioactive materials, plant operating limits, and specific
requirements for each SR SSC, including limiting conditions for
operation (LCO) and required surveillances. The proposed requirements
for TS would also include a section on important design elements, which
is similar to design features in Sec. 50.36, and a section for
administrative controls. A provision addressing the development and
submittal of TS to address decommissioning activities would also be
included in the proposed subpart G.
The proposed requirements for TS under part 53 would not carry over
safety limits or associated limiting safety system settings from 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 proposed construct of requirements for TS. In cases,
such as LWRs, where a SAFDL approach might be used as part of a
mechanistic approach to meeting the design and analysis requirements in
subpart C, the associated functional design criteria proposed in Sec.
53.410 and TS under the proposed Sec. 53.710(a) would define similar
requirements as those provided by the safety limit and limiting safety
system setting requirements in Sec. 50.36.
The proposed requirements for TS under part 53 would not include
specific criteria for identifying when LCOs must be established (i.e.,
would not include an equivalent to Sec. 50.36(c)(2)(ii)). Instead,
consistent with subparts B and C, the TS requirements in subpart F of
part 53 would define TS LCOs as providing limits on SR SSCs. The SR
SSCs protect against DBAs to demonstrate compliance with the safety
criteria in the proposed Sec. 53.210. In the proposed construct for
part 53, risk-significant SSCs would be addressed through a combination
of TS for the SR SSCs and establishment and monitoring of performance
standards for NSRSS SSCs.
In addition to addressing TS for SR SSCs, proposed Sec. 53.710
would require appropriate controls be developed and implemented for
NSRSS SSCs. Examples include appropriate surveillances and controls
established through reliability assurance programs. Configuration
management and other special treatments would provide that the
capabilities, availabilities, and reliabilities of NSRSS SSCs are
maintained consistent with the underlying risk assessments while
providing flexibility to licensees through maintaining the management
functions within licensee-controlled programs. Controls on NSRSS SSCs
are appropriate as part of the overall performance-based approach
within proposed part 53. Special treatments beyond those defined for
their SR functions may also be warranted for SR SSCs to reflect their
role in meeting the safety criteria in Sec. 53.220 and the evaluation
criteria in Sec. 53.450(e). The performance objectives for NSRSS SSCs
would reflect that the comprehensive risk metrics and related risk
performance objectives established under Sec. 53.220 may involve
assessing

[[Page 86934]]

and averaging the risks over a defined period (e.g., plant year) and
would not constitute a real-time requirement that must be continuously
demonstrated by the licensee. The controls under Sec. 53.710(b)
justify proposed changes in part 53 from the traditional or
deterministic approaches in parts 50 and 52 in areas such as replacing
the single-failure criterion with a probabilistic reliability criterion
(see SRM-SECY-03-0047, ``Policy Issues Related to Licensing Non-Light-
Water Reactor Designs,'' dated June 26, 2003). This approach could also
support the incorporation of risk insights and analytical margins to
gain operational flexibilities in areas such as siting and staffing
requirements described in subsequent sections of proposed subpart F.
Proposed Sec. 53.715 would provide the requirements for developing
and implementing a program to do the following: (1) control maintenance
activities; (2) take appropriate corrective action when performance
issues are identified; (3) conduct routine evaluations of
effectiveness; and (4) assess and manage risks resulting from
maintenance activities. These proposed requirements are similar to
those included in Sec. 50.65 (maintenance rule), including the need to
assess and manage the increase in risk that may result from the
proposed maintenance activities. While, for the maintenance rule,
specific criteria must be developed to capture both SR and non-SR but
otherwise important SSCs, the proposed Sec. 53.715 would cover SR SSCs
and NSRSS consistent with other subparts in part 53.
Proposed Sec. 53.720 would provide the requirements for responding
to a seismic event during the operating phase of the life cycle of a
commercial nuclear plant and would be equivalent to the requirements in
paragraph IV(a)(3) of appendix S, ``Earthquake Engineering Criteria for
Nuclear Power Plants,'' to part 50.
The proposed part 53 would include provisions to address staffing,
training, personnel qualifications, and human factors engineering (HFE)
in a manner that is risk informed, technology inclusive, performance
based, and flexible in nature. During the development of part 53, the
staff prepared a draft white paper on ``Risk Informed and Performance
Based Human-System Considerations for Advanced Reactors,'' to support
interactions with stakeholders and the ACRS. Key considerations include
the recognition that staffing, operator qualifications, and HFE are
interconnected areas that must be approached in an integrated manner
and, furthermore, that safety functions, including the means by which
they are fulfilled, provide an effective method for informing
technology-inclusive requirements.
The requirements associated with this approach would be in
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 would be introduced for
use within the context of subpart F. These new definitions would be the
following: ``Automation,'' ``Auxiliary operator,'' ``Generally licensed
reactor operator,'' ``Interaction-dependent-mitigation facility,''
``Load following,'' ``Self-reliant-mitigation facility.''
Sections 53.725 through 53.830 would be divided into four portions
that would cover general operational requirements, operator and senior
operator licensing requirements, generally licensed reactor operator
(GLRO) requirements, and general training requirements for plant staff.
The NRC intends to provide guidance addressing the review of operator
staffing plans; the review of operator, senior operator, and GLRO
examination programs; and the implementation of scalable HFE reviews.
Licensees would be required to use GLROs upon demonstrating compliance
with the criteria in Sec. 53.800.
Certain routine communications are necessary to facilitate the
operator licensing process. The NRC is proposing to adapt 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 is proposing to
adapt 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 is proposing to adapt 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 would provide 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 would
be 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 is proposing that these technical requirements would, together,
serve as a component of the required content of applications for OLs
and COLs under part 53. Additionally, the NRC proposes that the
specific technical requirements associated with HFE, human-system
interface design, concept of operations, functional requirements
analysis, and function allocation would serve as a component of the
required content of applications for standard DCs, standard design
approvals, MLs, and CPs, as well.
Human factors engineering is essential to facilitate the role of
personnel in facility safety in a manner that is both effective and
reliable. The NRC proposes to adapt Sec. 53.730(a) from the HFE design
requirements of Sec. 50.34(f)(2)(iii). A key difference would be that
the requirement would now be focused on settings where personnel
fulfill their safety or emergency response roles wherever they may
occur. The NRC additionally proposes to include within the scope of
this requirement activities for assuring the continued availability of
plant equipment that is needed for safety, and envisions that this may
encompass relevant maintenance, inspections, and testing as well. The
NRC intends that this requirement would be associated with staff
guidance for conducting scalable reviews of HFE that is planned to
accompany part 53.
Human-system interfaces provide vital information to operators
across a spectrum of operating conditions that can range from normal
operations through severe accident conditions. The specific types of
information that must be available to support operations staff during
such conditions include, in part, those associated with safety function
parameters, safety system status, possible core damage states, barrier
integrity, and radioactive leakage. Due to the importance of such
information, the NRC proposes under Sec. 53.730(b) to require such
human-system interface design features for all facilities, irrespective
of other flexibilities proposed under part 53. Therefore, the NRC
proposes to adapt specific post-Three Mile Island requirements of Sec.
50.34(f) in a technology-inclusive manner as detailed in the following:

[[Page 86935]]

<bullet> Paragraph (b)(1) would be adapted from Sec.
50.34(f)(2)(iv).
<bullet> Paragraph (b)(2) would be adapted from Sec.
50.34(f)(2)(v).
<bullet> Paragraph (b)(3) would be adapted from Sec.
50.34(f)(2)(xi), 50.34(f)(2)(xii), and 50.34(f)(2)(xxi).
<bullet> Paragraph (b)(4) would be 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) would be adapted from Sec.
50.34(f)(2)(xxvi).
<bullet> Paragraph (b)(6) would be 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 would be
provided under Sec. 53.730(b)(7). This prescriptive set of design
requirements for those facilities which demonstrate compliance with the
criteria of Sec. 53.800 would recognize that the application of HFE
under Sec. 53.730(a) is anticipated to be significantly reduced at
such facilities in the absence of an expected operator role for the
fulfillment of safety functions. However, it should be noted that the
capability for an immediately initiated, manual reactor shutdown would
be conservatively mandated irrespective of any other design
considerations.
The NRC proposes Sec. 53.730(c) to require the submittal of a
concept of operations that is of sufficient scope and detail to
appropriately inform the staff. The development of a concept of
operations can facilitate a clear understanding on the part of the NRC
for potential novel operating concepts. Additionally, such information
is likely to reduce the degree of resources and interactions needed for
the NRC to obtain the understanding necessary to enable flexible
requirements in areas such as staffing, operator qualifications, and
HFE.
The NRC proposes Sec. 53.730(d) to require the submittal of both a
Functional Requirements Analysis and a Function Allocation. The
identification of design-specific safety functions and how they are
fulfilled serves as a primary means for achieving technology-inclusive
requirements within areas such as staffing, operator qualifications,
and HFE. The Functional Requirements Analysis and Function Allocation
processes (which are both HFE methods derived from systems engineering
principles), provide an effective means to identify both how safety
functions will be satisfied and how to characterize any associated
operator role in doing so. A Functional Requirements Analysis shows
what features, systems, and human actions are relied upon to
demonstrate safety (i.e., fulfill safety functions). A Function
Allocation then describes how safety functions are assigned to both
personnel and automatic systems. However, an important adaptation of
the Function Allocation for use under the proposed rule would be the
further need to not only describe allocations of safety functions to
human action and automation, but also to identify allocations made to
active safety features, passive safety features, or inherent safety
characteristics as well.
Operating experience provides an important source of information by
which to inform various aspects of facility design and operations.
Accordingly, the NRC proposes in Sec. 53.730(e) to adapt 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 proposes under Sec.
53.730(f) to require the submittal of a staffing plan that details
operations staffing, how engineering expertise will be provided, and
what staffing will be available to provide other needed support
functions. The NRC intends that this requirement would be associated
with staff guidance for reviewing operations staffing plans that is
planned to accompany part 53 and that, following NRC approval of the OL
or COL, the staffing plan would become a condition of the facility
license. The NRC intends that, at a minimum, the approved licensed
operator and senior operator (or, if applicable, GLRO) staffing,
positions, and personnel locations will be incorporated into
corresponding requirements within the facility TS and that a license
amendment would thus be required for any subsequent changes.
Operator training and qualification programs provide an essential
component of supporting human performance in implementing tasks with
safety implications. Such programs must include components that cover
the stages of initial training, examination, and continuing training.
Additionally, recognizing the potential for varying concepts of
operations to affect traditional, prescriptive approaches to operator
proficiency, the NRC proposes under part 53 to allow facilities to
develop operator proficiency programs based on facility-specific
considerations.
Therefore, the NRC proposes in Sec. 53.730(g)(1) to require
approval as part of its approval of the OL or COL, of the programs that
will be used for the initial training, initial examination,
requalification training and examination, and proficiency of both
licensed operators and senior operators. In a corresponding manner, the
NRC proposes in Sec. 53.730(g)(2) to require approval of the programs
that will be used for the GLRO equivalents of each of these programs
for facilities with such staffing. The NRC intends that examination
program requirements would be associated with staff guidance for the
review of tailored examination processes that are planned to accompany
part 53. Following the completion of an initial training program,
continuing training programs provide an important means of sustaining
the knowledge and abilities of individuals. The NRC is proposing to
adapt the requirements of 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 proposes to
require these programs to be in effect concurrent with when the initial
operator examinations first commence, in effect putting the programs in
place only when they are needed. This represents a modification of the
comparable requirement of 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

[[Page 86936]]

an individual is participating in an approved training program or
reestablishing proficiency may also call for them to operate the
controls of the facility under the cognizance of a licensed individual.
The NRC is proposing to adapt the requirements of Sec. 55.13 in Sec.
53.735 to accomplish this, with a notable difference being the
incorporation of GLROs.
Section 53.740 would provide requirements for OL and COL holders
under part 53. Portions of Sec. 53.740 would be adapted from the
conditions of Sec. 50.54. In general, the conditions for operations
staffing under part 53 would reflect considerations for potential
technological differences and varying concepts of operation that are
expected among part 53 facility licensees. Additionally, certain
requirements would be specific to the operating phase while others
would remain in effect following the permanent cessation of facility
operations during the decommissioning phase.
All commercial nuclear plants licensed under part 53 would require
some form of licensed operator staffing, whether it be by specifically
or generally licensed operators. Consistent with this, the NRC is
proposing under Sec. 53.740(a) to require facility licensees to
demonstrate compliance with the programmatic requirements for either
specifically licensed operators and senior operators or for GLROs, as
applicable to the facility.
The NRC recognizes that 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 is
proposing under Sec. 53.740(b) to require that the staffing described
within the approved facility staffing plan be maintained as a condition
of the facility license as opposed to prescriptive staffing
requirements like those of 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 is proposing to adapt 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 is proposing to adapt 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 is therefore proposing to
adapt 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) would allow for load
following, provided that appropriate measures are in place. In
considering the acceptability of the measures associated with load
following, the NRC expects that any automatic protection relied upon
would be separate from that credited for reactor protection purposes
and would employ setpoints that are set so as to prevent actuation of
the reactor protection system while accomplishing its functions to the
extent practical.
Core alterations such as refueling are associated with specific
considerations that warrant limiting the oversight of such operations
to appropriately qualified and authorized individuals. Unlike other
types of fuel handling operations, core alterations occur within the
confines of a reactor vessel that is specifically designed to support
and sustain nuclear criticality, thereby justifying the imposition of
higher qualification levels within such contexts. The NRC is proposing
to adapt the requirements of 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
would already be required by either a specifically licensed senior
operator or by a GLRO, the NRC proposes to omit this requirement as
redundant during periods where core alterations occur while the plant
is operating.
It is impossible to predict every possible scenario that a
commercial nuclear plant might potentially encounter. Therefore, it is
prudent to grant the authority for appropriately qualified individuals
to depart from facility license conditions when emergency circumstances
dictate that doing so is in the interest of public health and safety.
The NRC is proposing to adapt the requirements of 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 proposes to extend this authority to

[[Page 86937]]

GLROs. While it is not anticipated that GLROs will have a role in the
fulfillment of safety functions at self-reliant-mitigation facilities
and, furthermore, that operators at such facilities would not be in a
position by which to significantly influence radiological safety
outcomes, the very nature of the Sec. 50.54(x) and (y) and the
proposed 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
Act authorizes the Commission to prescribe conditions for the licensing
of operators and to issue licenses consistent with those conditions.
The NRC is proposing to adapt the requirements of 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 proposes to license individuals as operators under both
specific and general licensing frameworks. Specific licenses would be
for licensed operators (i.e., reactor operators) and senior operators
(i.e., senior reactor operators) and would be issued to a named person
upon approval by the Commission of an application for that named
person. In contrast, GLROs would perform duties under the provisions of
a general license that would be effective without the filing of an
application with the Commission or the issuance of licensing documents
to a particular person. The NRC proposes requirements for the use of a
specific licensing process for licensed operators and senior operators
under 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 is proposing to
adapt 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 proposes to not extend a
comparable medical requirement to GLROs.
The NRC is also proposing to adapt 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 is
proposing to adapt 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 proposes to provide additional
flexibility by locating certain details associated with the preparation
and submittal of these applications within guidance in lieu of
placement within this proposed rule itself.
The NRC proposes overall programmatic requirements for specifically
licensed operator and senior operator training, examination, and
proficiency in Sec. 53.780. In general, the proposed 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
proposed 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 training program is based upon job requirements in a manner that
can be adapted to account for differences in plant technology, concepts
of operations, and operator roles in the fulfillment of design-specific
safety functions. The NRC is proposing under Sec. 53.780(a) to require
facility licensees to implement a SAT-based training program for the
initial training of licensed operator and senior operator applicants.
The program must be adequate to ensure that applicants will be capable
of performing the duties necessary both to protect public health and
safety and to maintain plant safety functions. The NRC further proposes
that such programs be subject to NRC approval and subsequent change
control processes of an appropriate nature.
Examinations provide a means of assessing that individuals have
achieved a degree of knowledge and ability that is sufficient to carry
out assigned duties as licensed operators or senior operators in a
manner that is safe and reliable. The NRC is proposing to adapt the
requirements of 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 would be that facilities have the flexibility
to propose, subject to NRC approval, the examination methods and
criteria to be used in assessing satisfactory applicant performance.
Such examination programs (including those used within the scope of
requalification training) would need to provide for acceptable levels
of both test validity and test reliability in order to be considered
acceptable. The NRC intends that staff guidance would be available to
facilitate the review of licensing examination programs that are
proposed by facility licensees and that, following NRC approval,
initial examination programs would be subject to an appropriate change
control process. Furthermore, the NRC proposes that holders of licenses
to operate commercial nuclear

[[Page 86938]]

plants under part 53 be provided the alternative of administering their
own approved licensing examinations. The NRC would continue to exercise
appropriate oversight of the program, make operator licensing decisions
based upon the examination results, and reserve the right to administer
the examinations in lieu of permitting the facility to do so. However,
irrespective of the provided flexibilities in examination format and
structure, at a minimum, topics from the following general categories
of knowledge and abilities should be sampled in such examinations:

<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 is proposing to adapt 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 would be that
distinct annual operating test and biennial written examination
components would not be mandated, with the facility licensee instead
proposing the examination methods and criteria to be used in assessing
satisfactory performance. The NRC intends that guidance would be
available to facilitate the review of the requalification examination
programs that are proposed by facility licensees and that, following
NRC approval, requalification examination programs would be subject to
an appropriate change control process.
For examinations to provide for valid assessments of the knowledge
and abilities of individuals, the examinations must remain free from
compromises that could affect their underlying integrity. The NRC is
proposing to adapt the requirements of 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 is proposing to adapt 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
proposed requirements of part 53 would not mandate that full scope,
plant-referenced simulators be used and would allow the use of
alternative simulation facilities consisting of, for example, partial
scope simulators or the plant itself, provided that all associated
requirements can be demonstrated to be met using alternative approaches
and methods. Additionally, in allowing for the possibility that an
applicant or licensee might demonstrate compliance with training,
examination, or experience requirements using the plant itself, the NRC
is not allowing the initiation of transients on the actual plant.
Consistent with this, aside from controlled reactivity manipulations
that are conducted for the purposes of demonstrating compliance with
experience requirements, actual plant components may not be operated
for these purposes. Rather, the NRC perspective is that the use of the
plant for training and examination purposes should be restricted to
techniques such as walkthroughs, job performance measures, simulated
tasks, use of augmented reality technology, and similar approaches that
provide training and examination value while avoiding the operation of
actual plant components.
There may be situations in which applicants for operator or senior
operator licenses have previous training and experience that justifies
waiving some, or all, of the initial examination requirements. The NRC
is proposing to adapt the requirements of 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 proposes to locate certain details associated with such waiver
requests within guidance documentation in lieu of placement within the
rule itself.
For licensed operators and senior operators to perform their
assigned duties safely and reliably, it is essential that they perform
those duties frequently enough so as to maintain a sufficient degree of
proficiency. The NRC is proposing to adapt the requirements of 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 is proposing, in contrast with the requirements of part 55, to
allow facility licensees to establish their own programs for operator
proficiency, subject to NRC approval.
As the holders of specific licenses, licensed operators and senior
operators must be subject to license conditions on an individual basis
to ensure that the basis upon which the licenses were issued remains
valid. The NRC is proposing to adapt the requirements of 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
proposing to allow certain aspects of operator proficiency to be
addressed by an NRC-approved facility proficiency program.
Licenses for specifically licensed operators and senior operators
are issued by the NRC and must remain subject to modification or
revocation. The NRC is proposing to adapt the requirements of
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 six years, after which they expire,
unless otherwise renewed. The NRC is proposing to adapt 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 proposed rule, the NRC has discussed with
stakeholders the considerations that might justify the omission of the
specifically licensed operators and senior operators. However, even for
an inherently safe reactor with autonomous operation features, certain
important administrative functions (e.g., compliance with TS,
operability determinations, NRC notifications, emergency declarations,
risk assessment, maintenance oversight, and radiological release limit
compliance) would still need to be accomplished by

[[Page 86939]]

appropriately qualified and authorized individuals. Additionally, the
NRC recognized that manual manipulations of facility reactivity
controls must only be performed by individuals who have been
appropriately licensed by the Commission. The NRC therefore proposes
under Sec. 53.800 to establish 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 would employ GLROs rather
than specifically licensed operators and senior operators. The GLRO
regulations offer enhanced flexibilities and targeted relaxations in a
manner that is commensurate with the modified role of such operators to
ensure the safe operation of the associated facilities. In contrast,
those facilities not meeting the criteria of Sec. 53.800 would instead
be considered interaction-dependent-mitigation facilities and would
require staffing by specifically licensed operators and senior
operators. The terminology used to designate these facility types
reflects differences in how operators are anticipated to need to
interact with their plant systems in mitigating events and achieving
safe outcomes; such systems may either need operators to interact with
them in some manner (i.e., be interaction-dependent) or may instead be
able to rely fully upon their own capabilities independent of operator
interaction (i.e., be self-reliant).
Generally licensed reactor operators would differ from specifically
licensed operators because the latter would be directly and
independently evaluated by the NRC as part of their licensing process.
This direct and independent evaluation remains appropriate when
operators may reasonably be expected to exert a significant influence
on public health and safety outcomes. Therefore, a key determinant as
to whether generally licensed reactor operators can be utilized in
facility staffing is the assessment of the operator's role in
maintaining and fulfilling safety functions at the facility, such as
through the performance of credited actions for the mitigation of plant
events.
The criteria proposed in Sec. 53.800 would 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> adequate defense in depth achieved without reliance on
human action.
It should be noted that those facilities not meeting the criteria
proposed in Sec. 53.800 would instead be classified as interaction-
dependent-mitigation facilities and would require staffing by
specifically licensed operators and senior operators instead.
Generally licensed reactor operators would perform duties under the
provisions of a general license that would be effective without the
filing of an application with the Commission or the issuance of
licensing documents to a particular person. The NRC proposes
requirements for the general licensing process for GLROs under
Sec. Sec. 53.805 through 53.820. The requirements for GLROs would
parallel those for senior operators in regard to their comparable
administrative responsibilities. Nonetheless, the requirements for
GLROs would be relaxed and incorporate greater flexibilities compared
to the requirements for specifically licensed operators in a manner
that is consistent with the GLRO's role in safety at self-reliant-
mitigation facilities.
In order to use GLROs in lieu of specifically licensed operators
and senior operators, a OL/COL applicant would need to demonstrate that
its proposed facility is a self-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
would be required to report the identities of all generally licensed
reactor operators to the NRC on an annual basis. Furthermore, a
facility licensee must ensure that the facility design and performance
continue to meet the technological criteria to be classified as a self-
reliant-mitigation facility (i.e., the criteria of 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 would be predicated on this assumption. The NRC therefore proposes
under Sec. 53.805 to establish requirements for facility licensees
that address issues such as these. Finally, the failure of a self-
reliant-mitigation facility to subsequently meet the criteria of Sec.
53.800 after the issuance of an OL or COL would constitute a reportable
event (i.e., an unanalyzed condition that significantly degrades plant
safety) under the provisions of Sec. 53.1630.
The NRC proposes the general license for GLROs under Sec. 53.810.
GLROs would be licensed as a class of individuals under the provision
of Sec. 53.810(a) and would 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 would retain the ability to suspend or prohibit
individuals from operating under the general license should such action
be warranted.
The NRC proposes overall programmatic requirements for GLRO
training, examination, and proficiency under Sec. 53.815. In general,
these proposed requirements are adapted from those of part 55 and
parallel those also proposed for specifically licensed senior operators
in Sec. 53.780. These requirements include increased flexibilities and
several targeted relaxations that reflect the limited role of GLROs in
facility safety. The requirements proposed under Sec. 53.815 cover, in
part, the initial training, initial examination, continuing training,
requalification examination, and proficiency of GLROs. Section 53.805
would require the facility licensee to develop, implement, and maintain
these programs. Section 53.810, in turn, would prescribe that the
requirements of Sec. 53.805 would need to be met as a requirement of
the general license. The implication of this structure is that the
facility licensee would need to implement these programs for training,
examination, and proficiency, and GLROs would need to participate in
these programs to demonstrate compliance with the requirements of the
general license.
The initial training process provides GLROs with the knowledge and
abilities needed to fulfill assigned duties as GLROs. The use of a SAT
serves to ensure that the training program is based upon job
requirements in a manner that can be adapted to account for differences
in plant technology and concepts of operations. The NRC is proposing
under Sec. 53.815(b) to require facility licensees to 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 proposes that such
programs would be subject to NRC approval, oversight, and appropriate
change control processes. The training program must ensure that

[[Page 86940]]

GLROs maintain the necessary knowledge, skills, and abilities.
Examinations provide a means of assessing that individuals have
achieved a degree of knowledge and ability that will be sufficient to
enable them to carry out assigned duties as GLROs in a manner that is
both safe and reliable. The NRC proposes to adapt the requirements of
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
would be that facility licensees would be afforded the flexibility to
propose, subject to NRC approval, the examination methods and criteria
to be used in assessing satisfactory individual performance. Such
examination programs (including those used within the scope of
continuing training) would need to provide for acceptable levels of
both test validity and test reliability in order to be considered
acceptable. The NRC intends that staff guidance would be available to
facilitate the review of initial examination programs that are proposed
by facility licensees and that approved initial examination programs
would be subject to an appropriate change control process. In contrast
with both the requirements of part 55 and the proposed requirements of
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 is proposing to adapt the requirements of Sec. 55.59
in Sec. 53.815(b) to require that facility licensees implement both a
SAT-based continuing training program and a requalification examination
program. However, a notable difference from the examinations required
under part 55 would be that distinct annual operating test and biennial
written examination components would not be mandated. The facility
licensee would instead propose examination methods and criteria to be
used in assessing satisfactory performance. Furthermore, unlike the
comparable requirements of part 55 and those proposed for specifically
licensed operators and senior operators, a biennial periodicity for
requalification examinations would not be prescribed. However, adequate
justification for the proposed periodicity of requalification
examinations would be required. The NRC intends that staff guidance
would be available to facilitate the review of the requalification
examination programs that are proposed by facility licensees. Approved
requalification examination programs would be subject to an appropriate
change control process.
For examinations to provide for valid assessments of the knowledge
and abilities of individuals, the examinations must remain free from
compromises that could affect their underlying integrity. The NRC is
proposing to adapt the requirements of 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 is proposing to adapt 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 would not
be mandated. The potential use of alternative simulation facilities
consisting of, for example, partial scope simulators or the plant
itself, would be allowed provided that all associated requirements
could be demonstrated to be met using alternative approaches and
methods. Additionally, in allowing for the possibility that an
applicant or licensee might demonstrate compliance with training and
examination requirements using the plant itself, the NRC is not
allowing the initiation of transients on the actual plant. Consistent
with this, aside from controlled reactivity manipulations that are
conducted for the purposes of demonstrating compliance with experience
requirements, actual plant components may not be operated for these
purposes. Rather, the use of the plant for training and examination
purposes should be restricted to techniques such as walkthroughs, job
performance measures, simulated tasks, use of augmented reality
technology, and similar approaches that provide training and
examination value while avoiding the operation of actual plant
components.
There may be situations in which GLROs have previous training and
experience that justifies waiving some, or all, of the initial
examination. Therefore, the NRC is proposing under Sec. 53.815(f) to
allow facility licensees to waive some, or all, portions of initial
examinations provided that such waivers are consistent with a program
that has been approved by the NRC.
For GLROs to safely and reliably perform their assigned duties, it
is essential that they perform those duties frequently enough so as to
maintain a sufficient degree of proficiency. However, the NRC
recognizes that facilities that utilize GLROs may have concepts of
operation that warrant unique proficiency considerations. Therefore,
the NRC is proposing in Sec. 53.815(g) to require that facility
licensees develop, implement, and maintain programs to maintain and
reestablish, if needed, the proficiency of GLROs. This could occur, for
example, if an individual's extended absence from watch standing has
rendered proficiency requirements unmet.
The general license should remain in effect for an individual only
while that individual remains employed in a position that may call for
the individual to manipulate the reactivity controls of the facility.
The NRC proposes under Sec. 53.820 to require that the general license
would cease to be applicable on an individual basis when an
individual's employment status becomes such that this is no longer the
case. However, the NRC recognizes that for some types of self-reliant-
mitigation facilities, very long periods may elapse between
circumstances that necessitate manual manipulation of reactivity
controls. Therefore, the general license remains in effect for an
individual as long as the individual's current position could
potentially require that individual to manipulate reactivity controls
at some point within the course of the individual's assigned job
duties.
The NWPA, section 306 (42 U.S.C. 10226) authorizes and directs the
NRC to, in part, issue regulations and guidance that address the
training and

[[Page 86941]]

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 is proposing under Sec. 53.830 to adapt, 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 proposed part 53; conversely, and
for the same reasons, additional categories of plant personnel may need
to be covered by part 53. The NRC also recognizes that the timeframe
prescribed in 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 the
proposed part 53 reflects greater flexibilities in personnel categories
and programmatic timeframes, while still retaining the requirement that
such training programs be based on SAT.
The NRC is proposing under Sec. 53.830 to require SAT-based
training programs with the timeframe for when such programs are
required being based upon when the associated personnel are needed to
support facility-specific needs. The training programs would cover the
training and qualification of plant personnel in the general categories
of supervisors, technicians, and other appropriate operating personnel.
The licensee would not be required to seek NRC approval of a training
program prior to usage. However, the licensee is required to
accommodate NRC inspection of the training program. The NRC intends to
develop guidance to facilitate the inspection of these training
programs but does not intend for such guidance to preclude the
potential for the training programs to be maintained by a separate,
NRC-approved accreditation process.
The proposed Sec. 53.845 would require programs to be developed,
implemented, and maintained to help ensure that design features and
human actions have the capabilities and reliabilities necessary to
demonstrate compliance with the safety criteria in subpart B throughout
the operating life of each commercial nuclear plant. The proposed
programmatic requirements in subpart F would also address areas such as
radiation protection needed to control routine effluents during normal
operations. The proposed Sec. Sec. 53.850 through 53.910 would require
programs to support specific activities needed to ensure the prevention
or mitigation of unplanned events or to support normal operations for
any reactor design. However, each holder of an OL or COL would be
required to assess whether additional programs are needed for the
specific reactor design and location of the commercial nuclear plant.
Licensees would be able to combine, separate, and otherwise organize
programs and related documents as appropriate for the technologies and
organizations associated with the commercial nuclear plant.
Proposed Sec. 53.850 would require a radiation protection program
associated with the requirements in subparts B and C for public doses
resulting from normal operations and the protection of plant workers.
The proposed requirements related to doses from normal operations,
including routine effluents, would be similar to those specified in
Sec. 50.36a, ``Technical specifications on effluents from nuclear
power reactors,'' and related requirements in standard TS for offsite
dose calculation manuals. While the proposed section would include
requirements that are technically and programmatically similar to part
50, proposed Sec. 53.850 would not include a requirement for effluent-
related TS as is required in Sec. 50.36a. A proposed requirement
similar to that found in the administrative controls section of TS for
operating reactors licensed under parts 50 and 52 would be included for
programmatic controls of solid wastes to complement the design
requirements in proposed Sec. 53.425.
Proposed Sec. 53.855 would require 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 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, proposed 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
proposed Sec. 53.855 are not used to demonstrate compliance with
subpart B and subpart C of this part. Rather, compliance with the
requirements in proposed Sec. 53.855 would provide reasonable
assurance that adequate protective measures can and will be taken to
protect public health and safety in the event of a radiological
emergency.
Proposed Sec. 53.860 would identify the applicable regulations for
part 53 applicants related to the programs for physical security,
cybersecurity, FFD, AA, and information security. These programs are
discussed in more detail in section V, ``Changes to Other Parts of 10
CFR,'' of this document.
Proposed Sec. 53.860(a) would establish the physical protection
program and present a graded approach to physical protection
requirements. If a licensee can meet the proposed criterion in

[[Page 86942]]

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

\2\ SECY-22-0072, ``Proposed Rule: Alternative Physical Security
Requirements for Advanced Reactors,'' dated August 2, 2022.
---------------------------------------------------------------------------

Proposed Sec. 53.860(b) would require licensees to establish,
implement, and maintain an FFD program under part 26. Section 53.860(c)
would require licensees to establish, implement, and maintain an AA
program in accordance with either Sec. 73.56 or proposed Sec. 73.120,
as appropriate. Section 53.860(d) would require licensees to establish,
implement, and maintain a cybersecurity program in accordance with
either Sec. 73.54 or proposed Sec. 73.110. Section 53.860(e) would
require 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.
Proposed Sec. 53.865 would establish requirements for quality
assurance and refer to appendix B of part 50 for the part 53
requirements for SR design features. Proposed requirements related to
evaluating and reporting changes to the quality assurance program would
be included in proposed subpart I and would be equivalent to those
found in Sec. 50.54.
The proposed Sec. 53.870 would require licensees to actively
assess possible degradation of SSCs from the effects of aging, fatigue,
and environmental conditions. The proposed inclusion of requirements
related to designing and monitoring for possible degradation mechanisms
reflects important lessons learned from the history of LWRs and the
likely introduction of new design features and materials in future
commercial nuclear plants. The allowable combinations of design
features, operating experience, testing, and monitoring during
operations would support performance-based approaches to the initial
licensing of new technologies. The proposed performance-based approach
to integrity assessment programs would also allow for the subsequent
consideration of operating experience and appropriate corrective
actions or allowable relaxations for ensuring that design features
comply with the proposed functional design criteria of Sec. Sec.
53.410 and 53.420. The proposed program would be based upon a
comprehensive and integrated evaluation of the aging and other
degradation mechanisms applicable to the design; identification of the
affected SSCs; the allowances provided in the design of the SSCs for
degradation; and schedules and procedures for determining if and at
what rate degradation is occurring, as well as its cause. Risk insights
could be used to prioritize the monitoring, evaluation, and management
of degradation based upon the importance of the SSC to safety and the
time frame for when the effects of degradation could be of concern.
Proposed Sec. 53.875 would establish requirements for a fire
protection program supporting operations similar to Sec. 50.48. The
proposed fire protection program during operations would work in
concert with specific fire protection requirements proposed in subpart
C for design and analyses and in proposed subpart E for construction
and manufacturing.
Proposed Sec. 53.880 would establish requirements for an inservice
inspection (ISI) and inservice testing (IST) program, which are
historically important activities conducted in accordance with ASME
codes and regulations in Sec. 50.55a. While the proposed part 53 would
not incorporate specific consensus codes and standards into the
regulations, Sec. 53.880 allows for the use of generally accepted
codes and standards. The proposed requirement for an ISI and IST
program would reinforce the need to develop monitoring programs to be
conducted during a plant's operations phase to complement the design
process and address inherent uncertainties. The NRC encourages the
continued use of consensus codes and standards supporting design,
testing, and inspections to support integrated and performance-based
approaches in demonstrating compliance with the proposed requirements
in part 53.
Proposed Sec. 53.910 would establish requirements for developing,
implementing, and maintaining procedures (e.g., operations and
emergency operating procedures) and guidelines (e.g., accident
management guidelines). The programmatic requirements for many of the
procedures listed in this proposed section would be similar to the
requirements found in the administrative controls section of TS for
plants licensed under parts 50 and 52. The proposed inclusion, where
appropriate, of accident management guidelines in these requirements is
intended to ensure that an integrated set of procedures and guidelines
would be established by licensees to ensure command and control across
the spectrum of possible event sequences. The proposed required
procedures would also include those needed to complement the design
requirements in proposed 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

The proposed subpart G would provide the regulatory requirements
for the decommissioning phase of the life cycle of a commercial nuclear
plant.

[[Page 86943]]

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

Subpart H--Licenses, Certifications, and Approvals

Proposed subpart H would provide requirements related to
applications under part 53 for NRC licenses, certifications, or
approvals for commercial nuclear plants.
Proposed subpart H would specify requirements applicable to all
part 53 applications as well as requirements specific to part 53
applications for LWAs, ESPs, standard design approvals, standard DCs,
MLs, CPs, OLs, and COLs. Proposed subpart H would be equivalent to and
include all existing licensing, certification, and approval processes
currently covered under parts 50 and 52, with the exception of the
process for early review of site suitability issues. Interactions with
external stakeholders during the development of the proposed rule did
not identify significant interest in or need for including the process
for early review of site suitability issues in part 53.
Much of the proposed subpart H regulatory text is identical to the
corresponding language in parts 50 and 52, with minor changes to
account for cross references in part 53, to make language technology
neutral, or to reflect the unique analytical approach in part 53. In
these instances, this preamble discussion will describe the language as
``equivalent'' to the existing corresponding requirement in part 50 or
part 52 and will describe any deviations, where applicable.
Because part 53 carries over the majority of the licensing options
from parts 50 and 52, there are several sections in proposed subpart H
that are similar to existing regulations in parts 50 and 52. Proposed
Sec. 53.1100 would address filing of applications for licenses,
certifications, or approvals under oath or affirmation and is
equivalent to Sec. 50.30. The proposed Sec. 53.1100 does not include
the current requirement in Sec. 50.30(a)(2) that the applicant
maintain the capability to generate additional copies, because it is
unnecessary in the age of electronic submissions. In addition, the
existing requirement on applications for OLs in Sec. 50.30(d) is
included in proposed Sec. 53.1124(g)(2), ``Relationship between
sections,'' covering OLs, rather than in proposed Sec. 53.1100.
Proposed Sec. 53.1101 would lay out activities requiring an NRC
license and is equivalent to Sec. 50.10(b). Proposed Sec. 53.1103
would address combining applications and is equivalent to Sec. Sec.
50.31, 50.52, and 52.8. Proposed Sec. 53.1103(b) would continue the
Commission's practice of combining multiple authorizations for a
facility under parts 30, 40, 50, 52, and 70 into one license based on
the Commission's authority under Section 161h. of the Act to combine
NRC licenses. Proposed Sec. 53.1106 would address elimination of
repetition and is equivalent to Sec. 50.32.
Proposed Sec. 53.1109 would provide general information
requirements for the content of applications submitted to the NRC under
part 53 and is equivalent to Sec. 50.33, with the exception of Sec.
50.33(f) on financial qualifications, which is covered in proposed
subpart J, and Sec. 50.33(h) on earliest and latest dates for
completion of construction, which is covered in Sec. 53.1306 of this
subpart. Each application would need to include information to address
the items in proposed Sec. 53.1109 as cited in the appropriate section
of this subpart for the application type.
One change from current requirements can be found in proposed Sec.
53.1109(i), which is not limited to electricity generation as it is
currently in part 50. Some prospective NRC applicants are considering
development of nuclear plants for other commercial ventures, such as
process heat generation or hydrogen production. In addition, Sec.
53.1109(j), which requires applications containing classified
information to separate that information from the unclassified
information in the application, refers to ``Restricted Data or
classified National Security Information'' instead of the term used in
the corresponding provision in Sec. 50.33(j), ``Restricted Data or
other defense information.'' This change was made to use the defined
term in part 95 rather than ``defense information'' as used in Sec.
50.33(j). The usage in Sec. 50.33(j) dates back to the Atomic Energy
Commission amend

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