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Proposed Rule2026-19568

Regulatory Enhancements for Reactor Licensing, Decommissioning, and Operational Oversight

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Published

September 24, 2026

Issuing agencies

Nuclear Regulatory Commission

Abstract

Consistent with Executive Order 14300, "Ordering the Reform of the Nuclear Regulatory Commission," the NRC is conducting a review and wholesale revision of its regulations. This proposed rule primarily aims to provide regulatory enhancements for reactor licensing, decommissioning, and operational oversight and is one effort in the NRC's activities to address the direction in section 5 of Executive Order 14300.

Full Text

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<title>Federal Register, Volume 91 Issue 184 (Thursday, September 24, 2026)</title>
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[Federal Register Volume 91, Number 184 (Thursday, September 24, 2026)]
[Proposed Rules]
[Pages 60702-60790]
From the Federal Register Online via the Government Publishing Office [<a href="http://www.gpo.gov">www.gpo.gov</a>]
[FR Doc No: 2026-19568]

[[Page 60701]]

Vol. 91

Thursday,

No. 184

September 24, 2026

Part II

Nuclear Regulatory Commission

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10 CFR Parts 20, 21, 50, et al.

Regulatory Enhancements for Reactor Licensing, Decommissioning, and
Operational Oversight; Proposed Rule

Federal Register / Vol. 91 , No. 184 / Thursday, September 24, 2026 /
Proposed Rules

[[Page 60702]]

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NUCLEAR REGULATORY COMMISSION

10 CFR Parts 20, 21, 50, 52, 53, 55, 70, 72, and 75

[PRM-50-110, PRM-50-116, NRC-2025-1138, NRC-2020-0036, NRC-2018-0201,
and NRC-2015-0028]
RIN 3150-AL45

Regulatory Enhancements for Reactor Licensing, Decommissioning,
and Operational Oversight

AGENCY: Nuclear Regulatory Commission.

ACTION: Proposed rule and draft regulatory guides; request for comment.

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SUMMARY: Consistent with Executive Order 14300, ``Ordering the Reform
of the Nuclear Regulatory Commission,'' the NRC is conducting a review
and wholesale revision of its regulations. This proposed rule primarily
aims to provide regulatory enhancements for reactor licensing,
decommissioning, and operational oversight and is one effort in the
NRC's activities to address the direction in section 5 of Executive
Order 14300.

DATES: Comments must be submitted electronically using <a href="https://www.regulations.gov">https://www.regulations.gov</a> by 11:59 p.m. eastern time on November 9, 2026.

ADDRESSES: Submit your comments, identified by Docket ID NRC-2025-1138,
at <a href="https://www.regulations.gov">https://www.regulations.gov</a>. If your material cannot be submitted
using <a href="https://www.regulations.gov">https://www.regulations.gov</a>, call or email the individual listed
in the FOR FURTHER INFORMATION CONTACT section of this document for
alternate instructions.
Do not include any personally identifiable information (such as
name, address, or other contact information) or confidential business
information that you do not want publicly disclosed. All comments are
public records; they are publicly displayed exactly as received, and
will not be deleted, modified, or redacted. Comments may be submitted
anonymously.
Follow the search instructions on <a href="https://www.regulations.gov">https://www.regulations.gov</a> to
view public comments.
You can read a plain language description of this proposed rule at
<a href="https://www.regulations.gov/docket/NRC-2025-1138">https://www.regulations.gov/docket/NRC-2025-1138</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: Aaron Kwok, Office of Nuclear Material
Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington,
DC 20555-0001; telephone: 301-415-1371, email: <a href="/cdn-cgi/l/email-protection#f8b9998a9796d6b38f9793b8968a9bd69f978e">[email&#160;protected]</a>.

SUPPLEMENTARY INFORMATION:

Executive Summary

A. Need for the Regulatory Action

On May 23, 2025, President Trump signed Executive Order (E.O.)
14300, ``Ordering the Reform of the Nuclear Regulatory Commission.''
E.O. 14300 requires the NRC to undertake a review and wholesale
revision of its regulations and guidance documents. In accordance with
E.O. 14300, the NRC identified potential changes across title 10 of the
Code of Federal Regulations (10 CFR) parts 20, 21, 50, 52, 53, 55, 70,
72, and 75. The NRC anticipates that these changes would yield
significant efficiencies and reduce regulatory burden while continuing
to provide reasonable assurance of adequate protection of the health
and safety of the public and common defense and security, and would
also ensure an appropriate level of environmental review.

B. Major Provisions

Major provisions of this proposed rule, supported by accompanying
draft guidance (if applicable), include changes in the following areas:

a. Part 21 Clarifications (10 CFR Parts 21, 50, and 53)
b. Reporting Requirements for Nonemergency Events for Nuclear Power
Plants (10 CFR Part 50)
c. Decommissioning Licensing Enhancements (10 CFR Parts 20, 50, 52,
and 53)
d. Financial Qualifications and Decommissioning Trust Fund Use (10
CFR Parts 50, 52, and 53)
e. Risk Informed Seismic Design (10 CFR Part 50)
f. Probabilistic Risk Assessment Requirements (10 CFR Part 50)
g. Backfitting and Issue Finality Provisions (10 CFR Parts 50, 52,
53, 70, and 72)
h. Notification of Initial Fuel Load (10 CFR Parts 52 and 53)
i. Requirements for Evaluation of Generic Issues and Operating
Experience (10 CFR Part 52)
j. Standard Design Approvals (10 CFR Parts 50, 52, and 53)
k. Environmental Reviews of Early Site Permits (10 CFR Parts 52 and
53)
l. Post-TMI Requirements (10 CFR Part 50)
m. Updates to the Operator Licensing Programs (10 CFR Parts 50, 53,
and 55)
n. International Atomic Energy Agency (IAEA) Notifications (10 CFR
Part 75)
o. Record Keeping and Reporting Requirements (10 CFR Parts 50, 52,
53, 70, and 72)

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 draft guidance as well as qualitative factors to be
considered in the NRC's rulemaking decision. The conclusion from the
analysis is that this proposed rule and associated draft guidance would
result in net averted costs (i.e., savings) to the industry and the
NRC, over the next 30 years, ranging from $311 million using a 7-
percent discount rate to $411 million using a 3-percent discount rate.
For the industry, the net averted cost is estimated at $305 million (7
percent net present value (NPV) and $398 million (3 percent NPV). For
the NRC, the net averted cost is estimated at $5.50 million (7 percent
NPV) and $13.9 million (3 percent NPV). On an annualized basis, the net
averted cost to the industry and the NRC would be about $22.2 million
per year at a 7 percent discount rate and $15.1 million per year at a 3
percent discount rate.
The draft regulatory analysis also considers qualitative factors,
such as regulatory efficiency. These benefits would result from
clarifications, administrative changes, and streamlining of processes
(such as notifications).
For more information, please see the draft regulatory analysis
(available in the NRC's Agencywide Documents Access and Management
System (ADAMS) Accession No. ML26191A071).

Table of Contents

I. Obtaining Information and Submitting Comments
II. Executive Order 14300: Ordering the Reform of the Nuclear
Regulatory Commission
III. Background
IV. Discussion
V. Specific Requests for Comments
VI. Regulatory Flexibility Certification
VII. Regulatory Analysis
VIII. Backfitting and Issue Finality
IX. Cumulative Effects of Regulation
X. Plain Writing
XI. National Environmental Policy Act
XII. Paperwork Reduction Act
XIII. Executive Orders
A. Executive Order 12866: Regulatory Planning and Review (as
Amended by Executive Order 14215, Ensuring Accountability for All
Agencies)
B. Executive Order 14154: Unleashing American Energy
C. Executive Order 14192: Unleashing Prosperity Through
Deregulation
D. Executive Order 14267: Reducing Anti-Competitive Regulatory
Barriers
E. Executive Order 14270: Zero-Based Regulatory Budgeting To
Unleash American Energy
XIV. Compatibility of Agreement State Regulations
XV. Availability of Guidance
XVI. Availability of Documents

[[Page 60703]]

I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID NRC-2025-1138 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-2025-1138.
<bullet> NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly available documents online in the
ADAMS Public Documents collection at <a href="https://www.nrc.gov/reading-rm/adams.html">https://www.nrc.gov/reading-rm/adams.html</a>. To begin the search, select ``Begin ADAMS Public 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#ca9a8e98e498afb9a5bfb8a9af8aa4b8a9e4ada5bc">[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#68382c3a463a0d1b071d1a0b0d28061a0b460f071e">[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.
<bullet> Public Meeting: The NRC may conduct a public meeting to
describe the proposed amendments and answer questions from the public
on this proposed rule. If the NRC determines it will hold a public
meeting, NRC will publish a notice of the location, time, and agenda of
the meeting on the NRC's public meeting website within 10 calendar days
of the meeting. Stakeholders should monitor the NRC's public meeting
website for information about the public meeting at: <a href="https://www.nrc.gov/public-involve/public-meetings/index.cfm">https://www.nrc.gov/public-involve/public-meetings/index.cfm</a>.

B. Submitting Comments

Comments must be submitted using <a href="https://www.regulations.gov">https://www.regulations.gov</a> by
11:59 p.m. eastern time on November 9, 2026. Please include Docket ID
NRC-2025-1138 in your comment submission.
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. Executive Order 14300: Ordering the Reform of the Nuclear
Regulatory Commission

On May 23, 2025, President Donald J. Trump signed E.O. 14300,
``Ordering the Reform of the Nuclear Regulatory Commission.'' Section
5, ``Reforming and Modernizing the NRC's Regulations,'' of that E.O.
requires the NRC to undertake a review and wholesale revision of its
regulations and guidance documents as guided by the policies set forth
in section 2 of the E.O. This rulemaking addresses the required review
and wholesale revision of the NRC's regulations and guidance documents
specified in the main body of section 5 of E.O. 14300.

III. Background

Over the decades, under its authority to make, promulgate, issue,
rescind, and amend such rules and regulations as may be necessary, the
NRC has developed a comprehensive and evolving regulatory framework to
ensure that licensee programs at nuclear facilities continue to provide
reasonable assurance of adequate protection to the health and safety of
the public and common defense and security. This proposed rule seeks to
modernize the NRC's regulations across 10 CFR parts 20, 21, 50, 52, 53,
55, 70, 72, and 75 with a focus on reducing or eliminating duplicative
or unnecessary requirements, providing flexibility, and implementing
risk-informed, performance-based alternatives where appropriate with
respect to reactor licensing, decommissioning, and operational
oversight.
The NRC recently issued a final rule establishing 10 CFR part 53,
``Risk-Informed, Technology-Inclusive Regulatory Framework for
Commercial Nuclear Power Plants'' (91 FR 15696, March 30, 2026). Part
53 sets forth a technology-inclusive regulatory framework as an
alternative to the NRC's current application and licensing
requirements, which were primarily developed to address license
requests concerning light water-cooled reactors and operational
requirements for those types of reactors. As discussed in Section IV of
this document, proposed changes would apply to licensees and applicants
under 10 CFR parts 50, 52, and 53, as applicable. The proposed changes
would provide enhanced regulatory flexibility for both current and
future licensees, streamline administrative processes, and ensure that
NRC requirements remain effective, efficient, and responsive to
credible risks.
The NRC prepared an unofficial redline strikeout version of the
proposed changes to regulatory text that is intended to help the reader
identify the proposed changes. The unofficial redline strikeout version
of this proposed rule is publicly available and is listed in the
``Availability of Documents'' section.
For the convenience of the reader, this proposed rule has been
organized as follows within each section:

a. 10 CFR Parts 21, 50, and 53--Part 21 Clarifications
b. 10 CFR Part 50--Reporting Requirements for Nonemergency Events
for Nuclear Power Plants
c. 10 CFR Parts 20, 50, 52, and 53--Decommissioning Licensing
Enhancements
d. 10 CFR Parts 50, 52, and 53--Financial Qualifications and
Decommissioning Trust Fund Use
e. 10 CFR Part 50--Risk Informed Seismic Design
f. 10 CFR Part 50--Probabilistic Risk Assessment Requirements
g. 10 CFR Parts 50, 52, 53, 70, and 72--Backfitting and Issue
Finality Provisions
h. 10 CFR Parts 52 and 53--Notification of Initial Fuel Load
i. 10 CFR Part 52--Requirements for Evaluation of Generic Issues and
Operating Experience
j. 10 CFR Parts 50, 52, and 53--Standard Design Approvals
k. 10 CFR Parts 52 and 53--Environmental Reviews of Early Site
Permits
l. 10 CFR Part 50--Post-TMI Requirements
m. 10 CFR Parts 50, 53, and 55--Updates to the Operator Licensing
Programs
n. 10 CFR Part 75--IAEA Notifications
o. 10 CFR Parts 50, 52, 53, 70, and 72--Record Keeping and Reporting
Requirements

10 CFR Parts 21, 50, and 53--Part 21 Clarifications

The NRC is proposing revisions to 10 CFR part 21, ``Reporting of
Defects and Noncompliance,'' for reporting safety defects. The
regulations in 10 CFR part 21 establish procedures and requirements for
the implementation of section 206 of the Energy Reorganization Act of
1974 (ERA). At a

[[Page 60704]]

high-level, 10 CFR part 21 addresses risks from basic components that
have been identified to contain a defect, which could result in a plant
failing to meet its licensing basis. The requirements and procedures of
10 CFR part 21 contribute to public health and safety by ensuring the
Commission is adequately informed of any loss of safety function to the
extent that there is a major reduction in the degree of protection
provided to the public health and safety. Under 10 CFR part 21,
directors and responsible officers of firms and organizations building,
operating, or owning NRC-licensed facilities are required to report
failures to comply with regulatory requirements relating to substantial
safety hazards and defects in components that may result in a
substantial safety hazard. This regulation also applies to directors
and responsible officers of firms and organizations supplying safety-
related components, as well as those providing safety-related design,
testing, inspection, and consulting services. Part 21 also requires
these entities to adopt procedures to ensure that safety-related
defects and noncompliance are brought to the attention of responsible
officers and directors (or their designees). In turn, the responsible
officers and directors (or their designees) are required to notify the
Commission by filing an initial report followed by a written report
regarding the defect or noncompliance.
For facilities licensed under 10 CFR parts 40, 70, and 21 has been
applicable since its publication in the final rule ``Reports to the
Commission Concerning Defects and Noncompliance'' (42 FR 28891; June 6,
1977). Since then, the lack of clarity in 10 CFR part 21 has presented
challenges to licensees, vendors, and the NRC staff. The rule's
preamble discusses the broad, overarching nature of the regulations in
10 CFR part 21. In particular, it identifies the need for additional
information to be accumulated after the rule has been implemented, with
a recommendation that additional guidance clarify the regulation once
experience has been gained. In response, the NRC published NUREG-0302,
``Remarks Presented (Questions/Answers Discussed) at Public Regional
Meetings to Discuss Regulations (10 CFR part 21) for Reporting of
Defects and Noncompliance,'' to provide further clarity, including
thresholds for the substantial safety hazard criterion of ``moderate
exposure to, or release of, licensed material.'' The information
presented in NUREG-0302 does not provide specific guidance to
applicants or licensees on how to comply with 10 CFR part 21; rather,
it is a collection of questions and answers from a series of public
meetings regarding 10 CFR part 21 requirements held from July 12-26,
1977. Additionally, the NRC staff has not provided detailed guidance to
clarify the nexus between 10 CFR parts 21 and 70 requirements including
the updated regulatory requirements in subpart H to 10 CFR part 70,
``Additional Requirements for Certain Licensees Authorized to Possess a
Critical Mass of Special Nuclear Material'' issued in September 2000,
such as which types of structures, systems or components at a facility
licensed under 10 CFR part 70 would be subject to the basic component
requirements in 10 CFR part 21.
The NRC first amended 10 CFR part 21 in 1978 in order to exempt
commercial-grade items (CGIs) from 10 CFR part 21 regulations until the
items were dedicated for use as basic components (43 FR 48621; October
19, 1978). This rulemaking established the definition of CGIs and the
commercial-grade dedication process, which would still be applicable to
reactor and nonreactor facilities. In 1991, the NRC issued Information
Notice 91-39, ``Compliance with 10 CFR part 21, `Reporting of Defects
and Noncompliance,' '' to inform NRC materials licensees of their
responsibilities under 10 CFR part 21. Additionally in 1991, 10 CFR
part 21 was amended to clarify the definitions of ``defect'' and
``basic component'' as well as to add the requirement for the retention
of 10 CFR part 21 records (56 FR 36081; July 31, 1991). This rulemaking
did not clarify the definition for, nor did it give specificity for the
identification of, nonreactor basic components. In 1995, the NRC
updated 10 CFR part 21 again (60 FR 48369; September 19, 1995) to
provide flexibility in procuring CGIs for safety-related equipment and
services to nuclear power plants; however, the issue of CGIs with
respect to nonpower reactor licensees was not changed.
In 2010 and 2011, the NRC's Office of the Inspector General (OIG)
performed two audits related to 10 CFR part 21. These audits resulted
in 15 recommendations, most of which related to clarifying 10 CFR part
21.\1\ One of the issues identified was that there was a lack of
clarity in the definition of ``basic component'' for nonreactor
facilities and activities that made it difficult to interpret the
applicability of the rule to 10 CFR part 70 licensees. This lack of
clarity was illustrated by the submittal of alternative definitions by
applicants and licensees to clarify the term ``basic component.'' The
NRC staff engaged stakeholders in multiple public meetings and also
undertook a series of site visits at fuel cycle facilities. These
interactions uncovered that there were widely differing interpretations
of the meaning of ``basic component'' at different fuel cycle
facilities. Specifically, at some facilities the term was interpreted
to include all systems, structures, and components designated as item
relied on for safety (IROFS) in accordance with Sec. 70.61(e), whereas
at other facilities the term was interpreted as applying only to those
IROFS that are the sole item relied on to prevent high consequence
events as defined in Sec. 70.61(b) or to mitigate their effects.
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\1\ OIG-10-A-20, ``Audit of NRC's Vendor Inspection Program,''
dated September 28, 2010 (ML102710583); OIG-11-A-08, ``Audit of
NRC's Implementation of 10 CFR part 21, Reporting of Defects and
Noncompliance,'' dated March 23, 2011 (ML110820426).
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In 2015, the NRC staff issued proposed changes to 10 CFR part 21
and an assessment of the areas that need clarification in ``Final
Regulatory Basis to Clarify 10 CFR part 21, `Reporting of Defects and
Noncompliance' '' (ML15152A457). This final regulatory basis proposed
clarifying the 10 CFR part 21 regulations through both rulemaking and
the development of regulatory guidance. However, these rulemaking
actions were subsequently discontinued (ML16028A189); therefore,
changes that would have clarified application of 10 CFR part 21 at fuel
facilities were not implemented.
As part of efforts to address E.O. 14300, the NRC has reassessed
whether certain potential changes identified in these prior activities
would be appropriate. This rulemaking proposes the following revisions
to 10 CFR part 21, paragraph (e) of Sec. 50.55, ``Conditions of
construction permits, early site permits, combined licenses, and
manufacturing licenses,'' and Sec. 53.605, ``Reporting of defects and
noncompliance,'' as a result of this reassessment and broader
assessment conducted pursuant to E.O. 14300. The proposed changes would
include consolidating specific requirements in Sec. Sec. 50.55(e) and
53.605 into 10 CFR part 21, allowing deletion of the redundant
regulations in Sec. Sec. 50.55(e) and 53.605.
The NRC proposes to amend five definitions: ``dedicating entity,''
``basic component,'' ``dedication,'' ``commercial grade item,'' and
``notification.'' One proposed change would revise the definition of
``dedicating entity'' to remove regulatory constraints that could limit
competition in accordance with E.O. 14267,

[[Page 60705]]

``Reducing Anti-Competitive Regulatory Barriers.'' This revision would
delete the description of who can perform the dedication; specifically,
it would remove references to a third-party dedicating entity. This
revision would also separate the definition of dedicating entity based
on the type of regulated facilities and the required quality assurance
program. Specifically, the definition has been separated into: (1)
facilities licensed under 10 CFR parts 50, 52, and 53; (2) 10 CFR part
70 licensees and other licensees, except for plutonium processing and
fuel fabrication plants or spent fuel reprocessing facilities, that are
required to develop an integrated safety analysis in accordance with 10
CFR part 70, subpart H; and (3) plutonium processing and fuel
fabrication plants or spent fuel reprocessing facility licensees
licensed pursuant to 10 CFR part 70 that are required to develop an
integrated safety analysis in accordance with subpart H to 10 CFR part
70. For the proposed definition of ``dedicating entity'' for plutonium
processing and fuel fabrication plants or spent fuel reprocessing
facilities, the definition would be revised to conform with a proposed
change to Sec. 70.22(f) \2\ to allow organizations with a dedication
process compliant with a 10 CFR part 50, appendix B or a proposed
appendix T, ``Streamlined Quality Assurance Criteria for Nuclear Power
Plants and Fuel Reprocessing Facilities,'' \3\ quality assurance
program to perform dedication activities. In addition, the definition
of ``dedicating entity'' would be modified to remove the reference to
Sec. 21.21(c) in order to not duplicate the requirements of Sec.
21.21(c). Another proposed revision would modify the definition of
``basic component'' to (1) provide clarity to fuel facility licensees
that are required to develop an integrated safety analysis to focus on
equipment that is a sole engineered IROFS and is essential to
preventing or mitigating an intermediate-consequence event and a high-
consequence event including an accidental criticality; (2) provide
clarity to plutonium processing and fuel fabrication plants or spent
fuel reprocessing facilities that are required to develop an integrated
safety analysis to focus on equipment that is relied on to prevent or
mitigate high-consequences to individuals located beyond the controlled
area boundary from a release of fission products and transuranics; (3)
provide consistency with the definition of safety-related SSCs under 10
CFR part 53; and (4) allow items designed and manufactured under a
quality assurance program that is complying with the proposed appendix
T, to 10 CFR part 50 to be included in the definition of a basic
component. The proposed changes to the ``dedication'' and ``commercial
grade item'' definitions would (1) clarify that nuclear material users
are exempt from the commercial grade dedication process; (2) add a
reference to 10 CFR part 52 to address concerns raised after the 2007
10 CFR part 52 rulemaking, in which the NRC recognized that conforming
changes to definitions in 10 CFR part 21 were not included to address
10 CFR part 52; (3) clarify that items that have successfully undergone
dedication are deemed equivalent to items designed and manufactured
under a quality assurance program that is complying with the proposed
appendix T; and (4) add a reference to 10 CFR part 53. The proposed
change to ``notification'' describes the primary method that the
communication to the NRC should occur.
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\2\ The proposed change to Sec. 70.22(f) is included in the
separate proposed rule, ``Modernizing Materials Licensing'' (91 FR
38214, June 24, 2026). The conforming changes proposed in this rule
to include a reference to 10 CFR part 50, appendix B in the
definition of ``dedicating entity'' in 10 CFR part 21 would,
accordingly, be contingent on the proposed addition of appendix B in
that proposed rule becoming effective.
\3\ The proposed addition of appendix T to 10 CFR part 50 is
included in the separate proposed rule, ``Modernizing Reactor
Licensing, Safety Oversight, and Siting Practices'' (91 FR 44560;
July 16, 2026). The conforming changes proposed in this rule to
include a reference to 10 CFR part 50, appendix T in the definition
of ``basic component'' in 10 CFR part 21 would, accordingly, be
contingent on the proposed addition of appendix T in that proposed
rule becoming effective.
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This proposed rule would remove 10 CFR part 76, ``Certification of
Gaseous Diffusion Plants,'' from the scope of 10 CFR part 21. This
proposed rule would also remove the ``Initial notification by
facsimile'' requirement in Sec. 21.21(d)(3)(i) and provide an updated
email submittal option. Overall, these proposed revisions would
increase clarity and reduce the regulatory burden.

10 CFR Part 50--Reporting Requirements for Nonemergency Events for
Nuclear Power Plants

The NRC is proposing to amend its regulations related to reporting
nonemergency events at nuclear power plants. The NRC would retain
reporting criteria that are important to plant safety and maintaining
appropriate NRC awareness of nonemergency events at nuclear power
plants. The NRC would eliminate criteria that are no longer needed.
This proposed rule would also allow for alternative reporting methods
for certain nonemergency events. The following subsections provide
further background on the specific reporting requirements that would be
revised or eliminated.
A. Immediate Notification Requirements Under Sec. 50.72
In a 1980 final rule, ``Immediate Reporting of Significant Events
at Operating Nuclear Power Reactors'' (45 FR 13434; February 29, 1980),
the NRC established in Sec. 50.72, ``Immediate notification
requirement for operating nuclear power reactors,'' immediate reporting
requirements for events considered significant. In the preamble to that
rule, the Commission stated that the capability of the NRC to make
timely decisions and to provide adequate assurances regarding actual or
potential threats to public health and safety, depends heavily on the
rapidity with which significant events are communicated by nuclear
power reactor licensees to the NRC. The majority of events occurring
throughout the nuclear industry pose little or no serious or immediate
threats to the public health and safety; however, certain events do
pose such threats or generate fear or unusual concern. The Commission
also stated that the NRC has an important obligation to collect facts
quickly and accurately about significant events, assess the facts, take
necessary action, and inform the public about the extent of the threat,
if any, to public health and safety. Not only must the NRC act promptly
to prevent or minimize possible injury to the public, but it must also
take appropriate action to alleviate fear or concern created as a
result of such events.
Accordingly, the NRC required licensees to report 12 types of
significant events to the NRC within 1 hour. These included serious
events that could result in an impact on public health and safety, such
as a nuclear power plant being in an uncontrolled condition, the
exceedance of a safety limit, or an uncontrolled release of
radioactivity.
Experience with the reporting requirements in Sec. 50.72 led to
the issuance of proposed amendments on December 21, 1981, ``Immediate
Notification Requirement for Operating Nuclear Reactors'' (46 FR
61894). Public comments on the proposed rule suggested that the
regulations should be revised to clarify reporting criteria and to
require early reports only on those matters of value to the exercise of
the NRC's responsibilities. The Commission amended the reporting
requirements of Sec. 50.72 in the 1983 final rule, ``Immediate
Notification Requirements of Significant Events at Operating Nuclear
Power Reactors'' (48 FR 39039;

[[Page 60706]]

August 29, 1983). The amendments clarified the list of reportable
events with the goal of providing the Commission with more useful
reports about the safety of operating nuclear power plants. The
Commission also created a category of ``nonemergency'' event reports.
Subsequently, the Commission amended the reporting requirements in
Sec. 50.72 in a final rule published on September 10, 1992 (57 FR
41378). The Commission determined that some of the event reports were
not necessary for the NRC to perform its safety mission and that
continued reporting of those events would not contribute useful
information to the operating reactor events database. The Commission
also considered that unnecessary reports would continue to consume
licensee and NRC resources that could be better applied elsewhere. As a
result, the NRC removed reports for events primarily involving invalid
engineered safety feature actuations that it determined were of little
or no safety significance.
In 2000, the Commission significantly modified the reporting
requirements of Sec. 50.72 in a final rule, ``Reporting Requirements
for Nuclear Power Reactors and Independent Spent Fuel Storage
Installations at Power Reactor Sites'' (65 FR 63769; October 25, 2000).
These amendments divided the nonemergency event reporting requirements
into 1-hour, 4-hour, and 8-hour notifications instead of only 1-hour
and 4-hour reporting requirements. The 1-hour reports were intended to
``capture those events where there may be a need for immediate action
by the NRC to protect public health and safety.'' For 4-hour reports,
the information was determined to be necessary in a prompt manner
because the NRC may need to respond to heightened public concern or
take a reasonably prompt action, such as partial activation of the
agency's response plan. The 8-hour reports were intended to capture
those events for which the NRC may need to take action within about a
day, such as initiation of a special inspection.
B. Recent Rulemaking Activity
On August 2, 2018, Bill Pitesa, on behalf of the Nuclear Energy
Institute (NEI), submitted a petition for rulemaking (PRM) under Sec.
2.802, ``Petition for rulemaking--requirements for filing,'' requesting
that the NRC amend Sec. 50.72 to remove all nonemergency event
notification requirements. In accordance with Sec. 2.803, ``Petition
for rulemaking--NRC action,'' the NRC docketed this petition for
rulemaking as PRM-50-116 (the petition) and reviewed the petition under
Sec. 2.803(h)(1). The NRC staff provided a rulemaking plan to the
Commission under SECY-20-0109, ``Petition for Rulemaking and Rulemaking
Plan on Immediate Notification Requirements for Nonemergency Events
(PRM-50-116; NRC-2018-0201),'' dated November 30, 2020, that
recommended evaluating the nonemergency event reporting requirements of
Sec. 50.72(b), ``Non-emergency events.'' In the staff requirements
memorandum (SRM) SRM-SECY-20-0109, ``Petition for Rulemaking and
Rulemaking Plan on Immediate Notification Requirements for Nonemergency
Events,'' dated July 28, 2021, the Commission directed the NRC staff to
proceed with a rulemaking on the immediate notification requirements
for nonemergency events in Sec. 50.72(b) and approved the closure of
the docket for PRM-50-116. The Commission provided additional direction
to the staff in SRM-SECY-20-0109 that (1) changes that shift the
responsibility for gathering, verifying, and communicating information
on nonemergency events from licensees to resident inspectors should not
be pursued, and (2) the staff should discontinue this effort if it
determines at any point that the costs are not justified by the
benefits or that the rulemaking would unacceptably degrade the NRC's
situational awareness.
Subsequently, the NRC published in the Federal Register a
regulatory basis to support rulemaking (87 FR 67571; November 9, 2022)
with a comment period ending on January 9, 2023. In the regulatory
basis, the NRC recommended maintaining five nonemergency event
reporting requirements, eliminating six nonemergency event reporting
requirements, revising guidance for two nonemergency event reporting
requirements, and adding an alternative reporting method. The NRC held
public meetings on November 4, 2021, on December 9, 2021, and on
December 13, 2022. The meeting summaries are available as indicated in
the ``Availability of Documents'' section of this document. The NRC
extended the public comment period for the regulatory basis to January
31, 2023 (87 FR 79258; December 27, 2022). The NRC received twelve
comment submissions on the regulatory basis, all of which were
considered in the development of this proposed rule and are available
in <a href="http://regulations.gov">regulations.gov</a> under Docket ID NRC-2020-0036.
C. NRC Observations on Stakeholder Feedback on the Regulatory Basis
The NRC received stakeholder feedback in several technical areas on
the regulatory basis. The public comment submissions are listed in the
following table, including associated ADAMS Accession Numbers. The NRC
compiled the annotated comment submissions into a single document,
which may be accessed within ADAMS using accession number ML23318A475.
The NRC reviewed these submissions, several of which resulted in
additional analysis and informed the development of this proposed rule.
The additional analysis did not result in changes to the
recommendations from the regulatory basis.
The NRC parsed each submission, depending on its length and
complexity, into one or more individual comments reflecting discrete
statements. The following discussion presents the NRC's summaries of
and responses to these comments.

ADAMS References for Public Comment Submissions on the Regulatory Basis
----------------------------------------------------------------------------------------------------------------
Submission No. Commenter Affiliation Accession No.
----------------------------------------------------------------------------------------------------------------
001................................ Stephen Monarque...... Private Citizen....... ML22363A563
002................................ Representative Kevin State of Connecticut ML23004A036
Ryan. Nuclear Energy
Advisory Council.
003................................ Rani Franovich........ The Breakthrough ML23011A311
Institute.
004................................ Phil Couture.......... Entergy Operations.... ML23025A364
005................................ David Gullott......... Constellation Energy ML23031A202
Generation.
006................................ Tony Brown............ Nuclear Energy ML23031A302
Institute.
007................................ Christopher P. Xcel Energy........... ML23031A303
Domingos.
008................................ Anthony Leshinskie.... State of Vermont, ML23034A224
Department of Public
Service.
009................................ Alyse Peterson........ New York State Energy ML23034A229
Research and
Development Authority.

[[Page 60707]]

010................................ Anonymous............. Unknown............... ML23034A230
011................................ Andrew Calcagno....... Private Citizen....... ML23034A231
012................................ Daniel Mills.......... Private Citizen....... ML23040A084
----------------------------------------------------------------------------------------------------------------

Summary of Comments and NRC Responses
Comment 1 (001-1, 003-1, 004-1, 005-1, 006-1, 007-1, 011-1): The
NRC received several comments in support of the proposed rulemaking and
encouraging the NRC to consider further changes to current event
notification requirements.
NRC Response: The NRC agrees, in part, with the comments. The NRC
is proposing changes consistent with the analysis the regulatory basis.
The NRC may consider additional changes to event notification
requirements in the future.
Accordingly, the Commission did not change the NRC staff's
recommendations in the regulatory basis in response to these comments.
Comment 2 (002-01, 002-02, 002-03, 008-1, 009-7, 009-8, 009-9, 012-
2, 012-3): Several comments reiterated the usefulness of event
notifications for the purposes of awareness and the ability for
stakeholders to consider taking actions independent of any NRC actions.
The proposed elimination of prompt nonemergency event notifications
would prevent the public from being adequately informed about many
issues at the plant until 60 days following an event, and the
methodology in the regulatory basis failed to analyze the importance of
prompt communications for effective responses to public inquiries.
Right now, the reports are made publicly available on the NRC website,
which is the timeliest source of information provided to the public.
The State, local, and Tribal governments are the most impacted, as
members of the public are more likely to reach out to trusted State and
local officials to understand the potential impacts of events. Absent
receiving these prompt public notifications, State and local officials
must expend valuable time and resources to reach out to licensees and
the NRC, thereby delaying response to the public.
NRC Response: The NRC agrees, in part, with the comments. Event
notifications may be useful for awareness of external parties; however,
the situational awareness of State and local officials for nonemergency
events is outside the scope of the NRC's regulatory authority. The NRC
is proposing to no longer compel event notification reports that are
not important to plant safety and are unnecessary to maintain
appropriate NRC awareness of nonemergency events at nuclear power
plants. The comments did not provide any examples of NRC situational
awareness that would be degraded in the absence of the nonemergency
event notifications recommended for elimination through the rulemaking.
Accordingly, the Commission did not change the bulk of the NRC
staff's recommendations in the regulatory basis in response to these
comments. The Commission did reconsider removal of reporting criterion
Sec. 50.72(b)(3)(xiii) for the major loss of offsite communications
capability, offsite response capability, and emergency assessment
capability. That reporting criterion is not proposed for removal.
Comment 3 (002-4): The NRC should engage further with communities
near commercial nuclear power plants for this rulemaking, including in-
person engagement, and perform an environmental justice impact review.
NRC Response: The NRC disagrees with the comment. Although the NRC
continuously aims to improve communications, the NRC determined that
the elimination of the recommended nonemergency event notifications
would not impact local communities because these notifications are not
needed for the NRC to fulfill its mission to provide reasonable
assurance of adequate protection of the public health and safety, which
includes the health and safety of communities around commercial nuclear
power plants. To that end, the NRC would retain a number of reporting
requirements, including the non-emergency event reports for (1)
deviations from technical specifications under Sec. 50.54(x) per Sec.
50.72(b)(1); (2) plant shutdowns required by technical specifications
per Sec. 50.72(b)(2)(i); (3) actuations of the reactor protection
system per Sec. 50.72(b)(2)(iv)(B); (4) the discovery of degraded or
unanalyzed conditions per Sec. 50.72(b)(3)(ii); and (5) the loss of
emergency response capabilities per Sec. 50.72(b)(xiii), as those
reporting requirements help facilitate the NRC's ability to ensure
public safety.
Accordingly, the Commission did not change the NRC staff's
recommendations in the regulatory basis in response to this comment.
Comment 4 (002-5, 008-2): The NRC should consider the costs to
States to create their own event notification requirements in lieu of
those the NRC recommends for deletion.
NRC Response: The NRC agrees with the comments. The NRC
incorporated changes to the regulatory analysis to account for States
that have indicated that they would initiate procedures to implement
any nonemergency event notification requirements at a State level that
would be eliminated by this proposed rule. The addition of these costs
did not change the overall cost-effectiveness of this proposed rule.
Accordingly, the Commission did not change the NRC staff's
recommendations in the regulatory basis in response to these comments.
Comment 5 (004-2): The NRC should be aware that the actual burden
hours of evaluating conditions for reportability may exceed the
estimated hours assumed in the cost analysis.
NRC Response: The NRC agrees with the comment. The hours assumed in
the regulatory analysis are the NRC's best estimate of average hours
expended per report. The NRC acknowledges that the actual burden of
reports is variable and that some reports may require more time than
assumed in the regulatory analysis. However, much of the burden
associated with evaluating the conditions would also be needed for
licensees' operability evaluations independent of reportability
requirements.
Accordingly, the Commission did not change the NRC staff's
recommendations in the regulatory basis in response to this comment.
Comment 6 (006-2, 006-03): The NRC should consider ways to
accelerate the rulemaking schedule and update existing guidance in
NUREG-1022 for reporting degraded or unanalyzed conditions under Sec.
50.72(b)(3)(ii) as soon as practical.
NRC Response: The NRC disagrees, in part, with the comment. The NRC
considered accelerating the schedule for this rulemaking but decided
not to

[[Page 60708]]

change the schedule. The rulemaking schedule provides time for proper
stakeholder engagement while ensuring that the rule and guidance
adequately and appropriately address all issues and public comments. If
the NRC proceeds to issue a final rule, the NRC anticipates publication
of a final rule in late 2026. Updates to the rulemaking schedule will
be made available in the NRC website: <a href="https://www.nrc.gov/reading-rm/doc-collections/rulemaking-ruleforum/active/ruleindex.html">https://www.nrc.gov/reading-rm/doc-collections/rulemaking-ruleforum/active/ruleindex.html</a>.
The NRC agrees that updates to existing guidance in NUREG-1022
solely for reporting degraded or unanalyzed conditions under Sec.
50.72(b)(3)(ii) can be achieved in advance of the rulemaking process.
The updated guidance would provide additional information to address
the high historical retraction rate of degraded or unanalyzed condition
notifications under Sec. 50.72(b)(3)(ii) in comparison to the other
nonemergency event reporting criteria. Because it is not proposing to
amend Sec. 50.72(b)(3)(ii), the NRC can revise the associated guidance
independent of this rulemaking effort and on a faster schedule than the
rulemaking. To that end, on October 2, 2024, the NRC published in the
Federal Register a notice of the issuance of NUREG-1022, Revision 3,
``Event Report Guidelines: Sec. Sec. 50.72 and 50.73,'' Supplement 2
(89 FR 80268), to provide clarifications for Section 3.2.4, ``Degraded
and Unanalyzed Conditions,'' of NUREG-1022, Revision 3.
Accordingly, the Commission did not change the NRC staff's
recommendations in the regulatory basis in response to these comments.
Comment 7 (008-3): It is unclear from the regulatory basis cost
estimates that the rulemaking will provide any significant reduction in
reporting burden.
NRC Response: The NRC disagrees with the comment. The regulatory
analysis performed for this rulemaking documents the reduction in
burden and associated costs resulting from this rulemaking. The
analysis shows that the rulemaking is generally cost effective.
Accordingly, the Commission did not change the NRC staff's
recommendations in the regulatory basis in response to this comment.
Comment 8 (009-1, 009-2):
The NRC should be careful not to reduce reporting requirements
further than prudent.
NRC Response: The NRC agrees with the comments. The regulatory
basis used a decision-making process driven by (1) past reporting data,
(2) NRC actions in response to those reports, and (3) a qualitative
risk assessment based on the NRC's Principles of Good Regulation and
the ``Be riskSMART'' framework, to determine which reporting
requirements must be maintained to support the NRC's ability to fulfill
its mission.
Accordingly, the Commission did not change the NRC staff's
recommendations in the regulatory basis in response to these comments.
Comment 9 (009-3): The NRC should consider additional reporting
requirements, such as those related to physical cybersecurity threats,
telecommunications incidents, or preparation for and response to severe
weather events.
NRC Response: The NRC disagrees with the comment. Although physical
and cybersecurity issues are not covered by the Sec. 50.72(b) event
reporting requirements, they are subject to reporting to the NRC via
other requirements such as those in Sec. 73.77, ``Cyber security event
notifications,'' and 73.1200, ``Notifications of physical security
events.'' Therefore, adding requirements to report these events under
Sec. 50.72(b) would be unnecessarily redundant.
Telecommunications issues impacting emergency communications are
currently reported to the NRC as required by Sec. 50.72(b)(3)(xiii).
Severe weather impacts are not specifically reportable to the NRC via
event notifications. However, the NRC maintains general awareness of
these impacts through other means, e.g., public meteorology sources,
routine inspections, and event reports required by conditions created
by the severe weather. Therefore, the NRC did not consider adding a
notification requirement for severe weather. Additionally, natural
phenomena, such as severe weather, that pose a threat to licensees are
reportable via licensee event report (LER) as required by Sec.
50.73(a)(2)(iii).
Accordingly, the Commission did not change the NRC staff's
recommendations in the regulatory basis in response to this comment.
Comment 10 (009-4): Eliminating the event notification reporting
requirements could result in licensees eliminating their procedures to
voluntarily notify resident inspectors of issues the same day. Also, it
would require resident inspectors to track due dates for licensee event
reports, identify when they are not provided within 60 days, and follow
up on non-reported items via the corrective action program, which is an
additional burden contrary to the Commission's directive.
NRC Response: The NRC disagrees with the comment. Although
licensees often notify resident inspectors of event notifications, any
internal licensee requirement to do so may be changed regardless of
this rulemaking, as there is no existing regulatory requirement for
them to notify resident inspectors. Resident inspectors have numerous
other means of maintaining awareness of plant activities and
conditions, which would not be impacted by this rulemaking.
Additionally, resident inspectors sample plant condition reports to
ensure licensees submit required notifications, including event
notifications and licensee event reports. This rulemaking would not
impact that activity.
Finally, resident inspector follow-up of events happens largely in
parallel with the development of event notification reports, not solely
in response to event notification reports. The removal of event
notifications would not substantially change the scope of resident
inspector activities. Rather, it means that plant conditions meeting
the criteria recommended for elimination under this rulemaking would no
longer require specific follow-up due to an event notification report.
Instead, such conditions would be treated as any other routine
condition of potential interest to resident inspectors, and any follow-
up activities inspectors choose to take would be driven via routine
baseline activities.
Accordingly, the Commission did not change the NRC staff's
recommendations in the regulatory basis in response to this comment.
Comment 11 (009-5, 009-6): The NRC should not eliminate both the 4-
hour reporting requirement to report discharge of the emergency core
cooling system (ECCS) in Sec. 50.72(b)(2)(iv)(A) and the requirement
to report safety system actuation in Sec. 50.72(b)(3)(iv), because
then there will be no notification of these types of events.
NRC Response: The NRC disagrees with the comments. In Section 14.0
of the regulatory basis, the NRC staff determined that the 4-hour event
notification requirement in Sec. 50.72(b)(2)(iv)(A) is unnecessary.
Nearly all events resulting in ECCS discharge into the reactor coolant
system that are reported to the NRC are captured by the Sec.
50.72(b)(2)(iv)(B) requirement to report reactor protection system
(RPS) actuations because most ECCS actuations are also RPS actuation
signals. The NRC's analysis determined that situations where an ECCS
actuates or is required to actuate, but a corresponding RPS actuation
is not required, are not significant enough to

[[Page 60709]]

warrant prompt NRC actions. Risk-significant ECCS actuations result in
corresponding RPS actuations when the reactor is at power. There are
few risk-significant conditions where ECCS actuations may occur and RPS
is not required to be active. Of those, such as during refueling
activities with the reactor coolant system partially drained, the
potential significance of the conditions is more appropriately captured
by licensees' emergency action levels (EALs) and the associated
notifications required in licensees' emergency plans. For non-risk-
significant conditions where an ECCS actuation occurs and RPS actuation
does not also occur, the NRC would not need to take prompt actions to
fulfill its mission. These conditions are appropriately captured by the
corresponding 60-day LER reporting requirements and routine baseline
inspection activities.
Separately, in Section 19.0 of the regulatory basis, the NRC
considered the 8-hour ECCS actuation reporting requirement as a subset
of the reporting requirements for general safety system actuations in
Sec. 50.72(b)(3)(iv). The NRC found that prompt NRC actions were not
warranted. For conditions reportable under this criterion (including
general ECCS actuation signals, but not ECCS discharges that also
result in RPS actuations), the NRC can fulfill its mission through
routine inspections and the corresponding 60-day LER reporting
requirements.
Accordingly, the Commission did not change the NRC staff's
recommendations in the regulatory basis in response to these comments.
Comment 12 (010-1): The NRC should make the conditions reportable
under Sec. 50.72(b)(3)(xiii) reportable as an LER.
NRC Response: The NRC disagrees with the comment. As discussed in
the ``Licensee Event Report System'' final rule (48 FR 33850, July 26,
1983), when the NRC established the LER system, the purpose of reports
required via Sec. 50.73 are for the NRC to obtain detailed narratives
of potentially safety significant events. In the review of past reports
made under this criterion, the NRC did not find any reports where the
conditions requiring an event notification under Sec.
50.72(b)(3)(xiii)--yet falling below the threshold of being reportable
under other criteria--would be of sufficient potential safety
significance to warrant the need for a detailed description through the
LER reporting of Sec. 50.73. As such, NRC awareness of these events
would be adequately maintained under the NRC's baseline inspection
program.
Accordingly, the Commission did not change the NRC staff's
recommendations in the regulatory basis in response to this comment.
Comment 13 (012-1): The proposed changes may increase the burden on
resident inspectors and require resident inspectors to devote more
inspection time to plant status.
NRC Response: The NRC disagrees with the comment. Resident
inspectors maintain awareness of a broad array of plant activities and
conditions. Most of these conditions are not reportable. The
elimination of the proposed reporting requirements does not imply that
residents should track those conditions outside of event reports.
Residents would still maintain awareness of plant activities and
conditions via their normal methods, and residents would not have to
give any more or less attention to those that were previously
reportable than they would give to other plant conditions of interest
that were not previously reportable.
Accordingly, the Commission did not change the NRC staff's
recommendations in the regulatory basis in response to this comment.
Comment 14 (012-4): There appears to be a gap in the current
requirements. The NRC requires licensees to report a loss of outbound
(licensee to NRC) communication capability but does not require
licensees to report a loss of inbound (NRC to licensee) communication
capability. Section 50.72 should be modified to require notification of
the NRC for any loss of communications capability (inbound or
outbound).
NRC Response: The NRC agrees, in part, with this comment. The
current reporting requirements do not require licensees to notify the
NRC of a loss of inbound communications capability. Licensees are not
required to report a loss of inbound communications because there is no
reliable way for licensees to test or otherwise know that inbound
communications capabilities are lost. However, the NRC tests inbound
communications daily with routine calls between licensees and the NRC's
Headquarters Operations Officers. This testing allows the NRC to become
aware of any loss of inbound communications capabilities.
Accordingly, the Commission did not change the NRC staff's
recommendations in the regulatory basis in response to this comment.

10 CFR Parts 20, 50, 52, and 53--Decommissioning Licensing Enhancements

The NRC is proposing revisions to decommissioning licensing
regulations in 10 CFR parts 20, 50, 52, and 53. Proposed changes to
decommissioning requirements build on prior NRC initiatives, including
the rulemaking effort discussed in SECY-24-0011, ``Regulatory
Improvements for Production and Utilization Facilities Transitioning to
Decommissioning.'' However, given that the Commission is presently
considering the NRC staff recommendations in SECY-24-0011, the
revisions in this proposed rule would be separate from, and would not
rely on, those recommendations.
A. Expansion of Partial Site Release Requests
In 2003, the NRC published a final rule that added Sec. 50.83,
``Release of part of a power reactor facility or site for unrestricted
use'' (68 FR 19711). Section 50.83 requires NRC approval for a partial
site release for unrestricted use at a reactor site before NRC approval
of the licensee's License Termination Plan (LTP). Before that final
rule, there was no guidance for allowing a portion of a site to be
released for unrestricted use, and the entire license footprint must
have been decommissioned at the end of the facility's operating period.
The flexibility of a partial site release allowed 10 CFR part 50 power
reactor licensees to request release of portions of the facility at any
time prior to the approval of the LTP, including during facility
operations. Partial releases following LTP approval would be governed
by the LTP or changes thereto, and partial releases for restricted use
are not permitted prior to LTP approval. Under Sec. 50.83, licensees
may request a partial site release by providing specific information to
the NRC, where the extent of the information necessary depends on
whether the area to be released has been designated as ``nonimpacted''
or ``impacted.''
In 2007, the NRC published a final rule amending 10 CFR part 52 (72
FR 49517). That final rule included in 10 CFR part 52 provisions
analogous to the decommissioning and license termination sections in 10
CFR part 50, yet the 10 CFR part 52 provisions reference only
Sec. Sec. 50.51 and 50.82, which do not include requests for partial
site release. Part 53 also includes a partial site release provision in
Sec. 53.1080, ``Release of part of a commercial nuclear plant or site
for unrestricted use.'' This proposed rule would correct this exclusion
in 10 CFR part 52 by including a new section, Sec. 52.111, ``Release
of part of a power reactor facility or site for unrestricted use,'' to
add a specific provision for release of part of a site for unrestricted
use before license termination for 10

[[Page 60710]]

CFR part 52 licensees. The language in proposed Sec. 52.111 is nearly
identical to the language in Sec. 50.83, with specific applicability
to holders of a combined license under 10 CFR part 52. The NRC is also
proposing minor conforming changes to 10 CFR part 20, ``Standards for
Protection Against Radiation,'' subpart E, ``Radiological Criteria for
License Termination,'' to include the release of part of a facility or
site under the proposed Sec. 52.111 and to limit the circumstances for
which the Commission could require additional cleanup following release
for unrestricted use. The proposed changes would provide 10 CFR part 52
licensees with a clear and efficient path toward releasing a portion of
their site prior to license termination and would align the regulations
in 10 CFR part 52 with those in 10 CFR parts 50 and 53, without any
reduction in safety or security.
B. Reducing Regulatory Requirements for Prompt Decommissioning of Power
Reactors
In 1996, the NRC amended its regulations to require power reactors
licensed under 10 CFR part 50 to submit a Post-Shutdown Decommissioning
Activities Report (PSDAR) before or within 2 years of permanent
cessation of operations instead of a detailed Decommissioning Plan (DP)
(61 FR 39278). Prior to the 1996 amendment, the DP approach, which is
currently the required licensing mechanism for nonpower production or
utilization facilities and all nuclear materials facilities, prohibited
power reactor licensees from performing dismantlement activities until
the NRC reviewed and approved the DP and issued an associated order
permitting decommissioning activities in accordance with the approved
plan.
Given that power reactor licensees have 60 years to terminate their
license per Sec. Sec. 50.82(a)(3), 52.110(c), and 53.1070(c), the
introduction of the PSDAR allowed these licensees to either begin
dismantlement activities promptly or place the reactor in storage and
delay dismantlement, the standard industry practice at the time. During
the development of the 1996 rule, the NRC assumed that power reactor
licensees would not be ready to terminate their license within a few
years after permanent cessation of operations. The PSDAR does not
require NRC approval, rather, under Sec. Sec. 50.82(a)(9)(i),
52.110(i)(1), and 53.1070(i)(1), the licensee must submit the LTP for
NRC review and approval before 2 years of expected license termination.
Decommissioning must be performed in accordance with the approved LTP.
Under current regulations in Sec. Sec. 50.82(a)(4)(i),
52.110(d)(1), and 53.1070(d)(1), the PSDAR submitted to the NRC must
contain a schedule of planned decommissioning activities, an estimate
of the decommissioning costs expected to be incurred, and a discussion
of environmental impacts of decommissioning. Under current regulations
in Sec. Sec. 50.82(a)(9)(ii), 52.110(i)(2), and 53.1070(i)(2), the LTP
submitted to the NRC must contain identification of remaining
dismantlement activities, an updated site-specific estimate of
remaining decommissioning costs, and supplement to the environmental
report describing any new information or significant environmental
change associated with the licensee's proposed termination activities.
In addition, the LTP must include a site characterization; plans for
site remediation; detailed plans for the final radiation survey; a
description of the end use of the site, if restricted; and
identification of parts, if any, of the facility or site that were
released for use before approval of the LTP. The NRC recognizes the
overlap in information submitted in both the PSDAR and LTP. If a
licensee planned to promptly terminate the license after permanent
shutdown, much of the information would be redundant.
The future landscape of power reactors licensed under 10 CFR parts
50, 52, and 53 will include various reactor technologies, types, and
sizes. Therefore, the decommissioning of such reactors will vary
tremendously, and accordingly, the NRC is proposing to proactively
amend the decommissioning licensing process to provide regulatory
enhancements and flexibility in the decommissioning and subsequent
license termination of all potential reactors. To accommodate licensees
who may decommission and terminate their license promptly after
shutdown, the NRC is proposing to eliminate the requirement to submit a
PSDAR if an LTP is submitted before or within 2 years of permanent
cessation of operations.
To implement this proposed change, the NRC proposes conforming
changes to certain requirements that commence at the time the PSDAR is
submitted in Sec. Sec. 50.82(a)(5), 50.82(a)(7), 52.110(e), 52.110(g),
53.1070(e), and 53.1070(g). The NRC is also proposing to amend the
duplicative requirement in Sec. Sec. 50.82(a)(8)(iii) and 52.110(h)(3)
related to submission of a site-specific decommissioning cost estimate.
C. Public Engagements Related to Decommissioning Licensing Actions
Under current NRC regulations Sec. Sec. 50.82(a)(4)(ii),
52.110(d)(2), and 53.1070(d)(2), the NRC must notice receipt of the
PSDAR in the Federal Register, make the PSDAR available for public
comment, hold a public meeting in the vicinity of the site being
decommissioned, and publish details about the public meeting in the
Federal Register and in a local forum. In the nearly 30 years since the
creation of the PSDAR, the NRC has gained experience with preparing and
reviewing PSDARs and conducting associated public interactions. Public
engagements related to submittal of a PSDAR impose significant costs on
the NRC and licensees without commensurate safety benefits, and public
comments have rarely resulted in changes to a licensee's PSDAR. This
proposed rule would eliminate the requirement for the NRC to notice
receipt of the PSDAR and make the PSDAR available for public comment.
The NRC would continue to use the information in the PSDAR to maintain
public health and safety at the beginning of the decommissioning
process.
Additionally, this proposed rule would modify the locality
requirement for the NRC to hold a public meeting for the PSDAR, LTP, or
a partial site release request ``in the vicinity of the site,'' as
required by Sec. Sec. 50.82(a)(4)(ii), 50.82(a)(9)(iii), 50.83(f),
52.110(d)(2), 52.110(i)(3), 52.111(f), 53.1070(d)(2), 53.1070(i)(3),
and 53.1080(f). This proposed rule would maintain the requirement for
the NRC to host a public meeting and modernize NRC regulations to
accommodate current and future practices for how the NRC currently
conducts public meetings in support of its long-standing position that
public involvement in NRC activities is a cornerstone of strong, fair
regulation of the nuclear industry.
D. Shortening the Period After Submittal of a PSDAR Before Major
Decommissioning Activities May Begin
Under current NRC regulations in Sec. Sec. 50.82(a)(5), 52.110(e),
and 53.1070(e), a power reactor licensee may not begin major
decommissioning activities until 90 days after the NRC receives the
PSDAR and until certifications of permanent cessation of operations and
permanent removal of fuel from the reactor vessel have been submitted.
The purpose of the current 90-day restriction was for the NRC to engage
stakeholders during its review of the PSDAR before dismantlement and
remediation began onsite. Because this proposed rule

[[Page 60711]]

would eliminate the required public comment period associated with the
PSDAR, the NRC's review of future PSDAR submittals would be completed
much faster. The NRC anticipates these reviews would be completed
within 30 days. Therefore, this proposed rule would reduce the
timeframe from 90 days to 30 days after the NRC receives the submittal
required before or within 2 years of permanent cessation of operations
(i.e., PSDAR or LTP).

10 CFR Parts 50, 52, and 53--Financial Qualifications and
Decommissioning Trust Fund Use

A. Conforming the Financial Qualifications Requirements for 10 CFR
Parts 50 and 52 Applicants With the Financial Qualifications
Requirements for 10 CFR Part 53 Applicants
Under Sec. Sec. 50.33(f) and 52.77, applications for utilization
facility permits and licenses under 10 CFR parts 50 and 52,
respectively, are required to include, except for an electric utility
applicant for a license to operate a utilization facility of the type
described in paragraph (b) of Sec. 50.21, ``Class 104 licenses; for
medical therapy and research and development facilities,'' or in Sec.
50.22, ``Class 103 licenses; for commercial and industrial
facilities,'' information sufficient to demonstrate that the applicant
``possesses or has reasonable assurance of obtaining'' the funds
necessary for construction and/or operation, as applicable, of the
facility along with associated financing details. However, for
commercial nuclear plants licensed under 10 CFR part 53, the NRC uses
the less restrictive financial qualifications standard of ``appears to
be financially qualified,'' which is similar to the standard used in
paragraph (a)(5) of Sec. 70.23, ``Requirements for the approval of
applications,'' for the possession of special nuclear material.
In 2018, the NRC staff submitted to the Commission a draft proposed
rule to change the NRC's financial qualifications requirements in 10
CFR part 50 to match the language of 10 CFR part 70. In 2022, the
Commission issued SRM-SECY-18-0026, ``Proposed Rule: Financial
Qualifications Requirements for Reactor Licensing,'' disapproving the
draft proposed rule. Instead, the Commission directed the staff to
address this issue during the development of 10 CFR part 53 and to
solicit stakeholder feedback on financial qualifications. As directed,
the staff solicited stakeholder feedback on using the 10 CFR part 70
standard in 10 CFR part 53 and, based on that feedback, 10 CFR part 53
ultimately did use the 10 CFR part 70 standard.
The NRC is now proposing to change Sec. 50.33(f)(1) and (2) so
that they conform to the financial qualifications standard for 10 CFR
part 53 applicants. Specifically, the proposed changes to Sec.
50.33(f)(1) and (2) would replace the current ``possesses or has
reasonable assurance of obtaining the funds necessary'' financial
qualifications standard with the 10 CFR part 53 ``appears to be
financially qualified'' financial qualifications standard. And, as with
10 CFR part 53, the proposed changes would provide that if available
funding at the time of application is 50 percent or less, the applicant
should include proposed license conditions to facilitate verification
that funding is available prior to the start of construction and/or
operation, as applicable. These changes would ensure that the NRC's
regulations are consistent.
B. Decommissioning Trust Fund Use
In 2024, the NRC issued ``Interim Staff Guidance: Use of the
Decommissioning Trust Fund During Operations for Major Radioactive
Component Disposal'' (89 FR 63447), providing both the NRC's position
on the topic and information that would assist the NRC staff in
evaluating a request for an exemption from Sec. 50.82 to allow for a
decommissioning trust fund (DTF) withdrawal for major radioactive
component disposal during operations. To date, the NRC has evaluated
and approved two separate exemption requests related to the use of the
DTF during operations and is currently evaluating a third. The NRC
expects a significant number of similar requests in the foreseeable
future. The NRC evaluates each exemption request for a DTF withdrawal
during operations to ensure that it meets the criteria for exemptions
in Sec. 50.12, ``Specific exemptions,'' and considers the totality of
facts in determining whether to grant or deny the request. Each request
is accompanied by a safety and environmental review, which results in
an average review time of 9 months.
To reduce time and resource burdens on both the industry and the
NRC, rather than relying on the exemption process, the NRC proposes to
establish a 60-working day notice/negative consent process. This
proposed rule changes would explicitly allow the use of DTFs during
operations for radiological decommissioning activities, including, but
not limited to, major radioactive component disposal, under certain
defined circumstances. This proposed rule would not change the NRC's
existing practice of evaluating a DTF withdrawal notice for
``extraordinary circumstances,'' using guidance similar to the current
Interim Staff Guidance. Additionally, proposed changes to Sec. Sec.
50.75 and 53.1045 would maintain the condition that licensees must
continue to have sufficient funds available for decommissioning in the
DTF. More information is provided in the draft Regulatory Guide (DG)-
1476 (proposed RG 1.159, Revision 4), ``Assuring the Availability of
Funds for Decommissioning Nuclear Reactors.''
The NRC would continue to implement its robust decommissioning
funding assurance oversight program, including the review of
decommissioning funding status reports for operating reactors. These
reports enable the NRC to ensure that funding will be available to
safely decommission a facility that has permanently ceased operation,
and to identify and respond to shortfalls in reasonable assurance,
should they occur. These activities would provide an appropriate level
of oversight to prevent funding shortages.

10 CFR Part 50--Risk Informed Seismic Design

Within the current regulatory framework, all power reactor
applicants under 10 CFR parts 50 and 52 are required to meet appendix
S, ``Earthquake Engineering Criteria for Nuclear Power Plants,'' to 10
CFR part 50, which does not allow for a graded approach to seismic
design based on the risks of the facility or the SSCs within the
facility. Appendix S to 10 CFR part 50 currently only allows for use of
a single design-basis ground motion (DBGM), referred to as the safe-
shutdown earthquake ground motion (SSE), and imposes a minimum peak
ground acceleration (consistent with the highest risk level) for the
response spectrum at the foundation level of structures.
Recently developed guidance, DG-4036 (proposed RG 4.29), ``Graded
Approach to Site Characterization for New Reactor Applications,'' which
would support changes to 10 CFR part 100, ``Reactor Site Criteria,'' in
the proposed rule, ``Modernizing Reactor Licensing, Safety Oversight,
and Siting Practices'' (91 FR 44560), would provide guidance on
determining appropriate DBGMs for a facility based on the potential
consequences from a design basis seismic event. DG-4036 also would
include an appendix providing guidance on optimization of a facility's
seismic design basis based on the use of multiple seismic design
categories and associated DBGMs. This proposed rule would modify
appendix S to 10 CFR part 50 to allow use of these risk-informed
approaches under 10 CFR

[[Page 60712]]

parts 50 and 52. A guidance document supporting this proposed rule, DG-
1471 (Draft RG 1.208 Revision 2), ``A Performance-Based Approach to
Define the Site-Specific Earthquake Ground Motion,'' would rely on the
proposed changes to 10 CFR part 100.
Section 53.480 includes a risk-informed and performance-based
alternative to appendix S to 10 CFR part 50 that allows for the use of
multiple DBGMs for 10 CFR part 53 applicants and licensees. The
different target performance goals associated with each DBGM provide
varying levels of reliability during a seismic event, which can be
assigned based on the safety-significance of SSCs or potential
radiological consequence resulting from failure of the SSCs during a
seismic event and enable the gradation of seismic design criteria for
SSCs. Similarly, and in a separate rulemaking, the proposed 10 CFR part
57, ``Licensing Requirements for Microreactors and Other Reactors with
Comparable Risk Profiles'' (91 FR 23628), would provide a risk-informed
and performance-based regulatory framework that would not preclude the
use of multiple DBGMs for 10 CFR part 57 applicants and licensees. The
proposed update to appendix S to 10 CFR part 50 in this rulemaking
would be consistent with the risk-informed and performance-based
approaches addressed in these other regulatory frameworks and DG-4036
and would enable the implementation of such approaches under 10 CFR
parts 50 and 52.
These proposed rules and guidance would enable nuclear facilities
to avoid overly conservative seismic design requirements by risk
informing seismic design to the appropriate level needed to ensure
safety.

10 CFR Part 50--Probabilistic Risk Assessment Requirements

Section 50.69, ``Risk-informed categorization and treatment of
structures, systems and components for nuclear power reactors,'' of 10
CFR provides an alternative, risk-informed approach for establishing
the requirements for the treatment of SSCs in nuclear power reactors.
Under this approach, SSCs are categorized according to their safety
significance. The current regulations permit this alternative approach
for holders of an operating license (OL) for a light water reactor
(LWR) under 10 CFR part 50, holders of a renewed LWR license under 10
CFR part 54, ``Requirements for Renewal of Operating Licenses for
Nuclear Power Plants,'' applicants for a construction permit (CP) or OL
under 10 CFR part 50, and applicants for a standard design approval
(SDA), a combined license (COL), or a manufacturing license (ML) under
10 CFR part 52.
The NRC is proposing to amend Sec. 50.69 to extend the
applicability of its alternative approach to holders of an LWR CP under
10 CFR part 50, as well as to applicants for an LWR design
certification and holders of an LWR COL and ML under 10 CFR part 52.

10 CFR parts 50, 52, 53, 70, and 72--Backfitting and Issue Finality
Provisions

The NRC regulations include several backfitting and issue finality
rules that provide for the predictability and stability of NRC
regulations. Although these rules vary to an extent, each generally
requires the NRC to demonstrate that specific criteria are met before
imposing new or revised requirements on certain holders of, and certain
applicants for, NRC approvals.
The NRC has always considered nuclear power reactors licensed under
section 103 or 104b of the Atomic Energy Act of 1954, as amended (AEA)
(i.e., commercial nuclear power reactors) to be within the scope of
Sec. 50.109, ``Backfitting.'' In 2026 (91 FR 2287), the NRC issued a
notice of interpretation explaining that commercial non-power
production or utilization facilities (NPUFs) licensed under section 103
of the AEA are within the scope of Sec. 50.109 and non-commercial
NPUFs licensed under section 104 of the AEA are excluded from the scope
of Sec. 50.109. This proposed rule would revise Sec. 50.109 to
clarify this applicability.
The current Sec. 50.109 does not require the NRC to consider the
costs of its actions in cases of ensuring, defining, or redefining
adequate protection, or in cases of ensuring compliance with NRC
requirements or conformance with written licensee commitments. Unless a
backfit is necessary for ensuring, defining, or redefining adequate
protection or is necessary for compliance, Sec. 50.109 requires the
NRC to demonstrate through a systematic and documented analysis (i.e.,
a backfit analysis) that there is a substantial increase in the overall
protection of public health and safety or the common defense and
security to be derived from the backfit and that the direct and
indirect costs of implementation are justified in view of this
increased protection.
The NRC also has backfitting rules in Sec. 70.76, ``Backfitting,''
for entities licensed to possess special nuclear material in quantities
greater than a critical mass and engage in specific activities (e.g.,
fuel facilities), and Sec. 72.62, ``Backfitting,'' for independent
spent fuel storage installations and monitored retrievable storage
installations. These regulations provide definitions of ``backfitting''
and related requirements similar to those in Sec. 50.109.
The regulations in 10 CFR part 52 include several issue finality
provisions related to the licensing of nuclear power reactors involving
early site permits, standard design certifications, COLs, SDAs, and
MLs. The design certification rules in the appendices to 10 CFR part 52
also include specific backfitting and issue finality provisions. The
issue finality provisions generally accomplish the same objective as
the backfitting provisions by limiting changes to certain decisions
made as part of the 10 CFR part 52 licensing process. For example,
there are several issue finality provisions in 10 CFR part 52 that
require justification similar to Sec. 50.109 for NRC-initiated changes
affecting issue finality.
The regulations in 10 CFR part 53 also include a backfitting rule
in Sec. 53.1590, ``Backfitting,'' and issue finality provisions
related to the licensing of nuclear power reactors under 10 CFR part
53. The backfitting rule and issue finality provisions in 10 CFR part
53 are generally equivalent to the backfitting rule in Sec. 50.109 and
issue finality provisions in 10 CFR part 52.
Backfitting generally does not include changes made by the NRC at
the request of an applicant, licensee, or similar entity, such as
changes made in response to a request for a license amendment,
exemption, or departure. Changes affecting issue finality can be
initiated by an applicant, licensee, or the NRC.
On November 29, 2016, the Commission issued SRM-COMSECY-16-0020,
``Revision of Guidance Concerning Consideration of Cost and
Applicability of Compliance Exception to Backfit Rule'' (ML16334A462),
and directed the staff to consider costs in compliance backfit
justifications and to make other changes in how the backfit and issue
finality rules are implemented. The term ``compliance backfit''
encompasses both backfitting actions and changes affecting issue
finality to ensure compliance with NRC requirements. This direction was
based, in part, on the Supreme Court's 2015 decision in Michigan v.
Environmental Protection Agency, 576 U.S. 743 (2105), that reflects the
view that under the Administrative Procedure Act, unless Congress has
indicated otherwise, an agency's decision-making calculus should
include at least some consideration of the cost placed on a

[[Page 60713]]

licensee to comply with new requirements.
Under the NRC's current regulations, the NRC uses a compliance
backfit justification when a modification is necessary to bring a
facility into compliance with NRC requirements or into conformance with
written commitments by the licensee. Compliance backfitting is limited
to addressing situations in which the NRC approved a method of
compliance with a requirement and later determines that (1) the method
does not comply with the requirement and (2) the previous approval
occurred because of an omission or mistake of fact at the time of
approval. New or modified interpretations of what constitutes
compliance are not within the scope of the compliance backfit
justification but may be considered under the other applicable
justifications for backfitting or imposing a change affecting issue
finality. Prior to SRM-COMSECY-16-0020, the NRC evaluation of
compliance and adequate protection backfits or changes affecting issue
finality were similar, and costs were not considered in either case.
On September 19, 2019, NRC Management Directive (MD) 8.4,
``Management of Backfitting, Forward Fitting, Issue Finality, and
Information Requests,'' was revised to reflect the Commission's updated
policy in SRM-COMSECY-16-0020 on the implementation of the backfitting
and issue finality rules. As described in the revised MD 8.4, if the
NRC is considering a backfitting action or change affecting issue
finality, it must first consider whether regulatory action is an issue
of adequate protection (i.e., necessary to provide reasonable assurance
of adequate protection or involving defining or redefining adequate
protection) and, if so, whether there is an imminent threat to public
health and safety. If regulatory action is not an issue of adequate
protection, then the NRC needs to determine if the proposed action
satisfies the applicable compliance backfit criteria, which includes
cost considerations. Finally, if the proposed action does not satisfy
the applicable compliance backfit criteria, then the NRC may consider
whether the proposed action is a cost-justified substantial increase in
overall protection. The issue finality provisions can vary from this
approach because some issue finality provisions allow for other
justifications and some do not allow for a cost-justified substantial
increase in overall protection.
Therefore, by the process described in MD 8.4, a compliance backfit
is not necessary for reasonable assurance of adequate protection of
public health and safety or the common defense and security. The NRC
must consider costs in the compliance backfit justification but, in
contrast to a cost-justified substantial increase in overall
protection, the NRC is afforded substantial flexibility in its cost
considerations and does not need to demonstrate a substantial increase
in overall protection.
On June 6, 2022, the NRC staff requested Commission approval to
publish the proposed rule described in SECY-22-0052, ``Proposed Rule:
Alignment of Licensing Processes and Lessons Learned from New Reactor
Licensing (RIN 3150-Al66)'' (package ML21159A055). In response, the
Commission issued SRM-SECY-22-0052 (ML24326A003), which approved the
publication of a revised proposed rule that would, among other things,
clarify the backfitting and issue finality requirements in 10 CFR parts
50 and 52 that govern changes to early site permits, SDAs, and
manufacturing licenses. These proposed clarifications and other
proposed changes are included in this rulemaking with some differences
to what was described in SECY-22-0052.
E.O. 14300, section 5(f), directs the NRC to ``[e]stablish
stringent thresholds for circumstances in which the NRC may demand
changes to reactor design once construction is underway'' as part of
the wholesale revision of its rules and guidance documents. The NRC's
backfitting regulations in Sec. Sec. 50.109 and 53.1590 and issue
finality provisions in 10 CFR parts 52 and 53, among other regulations,
include such thresholds. Additionally, in response to E.O. 14267, the
Commission identified that several backfitting and issue finality
provisions could be modified to reduce their competitive burden. In
response to E.O.s 14300 and 14267, the NRC proposes to amend its
backfitting and issue finality rules to eliminate the compliance
backfit exception, revise and clarify the applicability of the
backfitting and issue finality requirements, add criteria to allow the
NRC to reduce regulatory burden, and make other corrections.
Specifically, the proposed amendments would revise the backfitting
rules in Sec. Sec. 50.109, 53.1590, 70.76, and 72.62. This proposed
rule does not apply to the backfitting rule in 10 CFR part 76,
``Certification of Gaseous Diffusion Plants,'' because 10 CFR part 76
is no longer used and is scheduled to be sunset on January 8, 2027. The
proposed amendments would also revise the following issue finality
rules that apply to early site permits, design certifications, and
manufacturing licenses issued under 10 CFR parts 52 or 53: Sec. Sec.
52.31, ``Criteria for renewal,'' 52.39, ``Finality of early site permit
determinations,'' 52.59, ``Criteria for renewal,'' 52.63, ``Finality of
standard design certifications,'' 52.171, ``Finality of manufacturing
licenses; information requests,'' 52.179, ``Criteria for renewal,''
53.1176, ``Criteria for renewal,'' 53.1188, ``Finality of early site
permit determinations,'' 53.1257, ``Criteria for renewal,'' 53.1263,
``Finality of standard design certifications,'' 53.1288, ``Finality of
manufacturing licenses,'' and 53.1295, ``Renewal of manufacturing
licenses.''
In addition, the proposed amendments would revise the plant-
specific backfitting limitations in paragraph VIII.B.3 of the five
design certification rules currently in effect:
a. Appendix A to 10 CFR part 52, ``Design Certification Rule for
the U.S. Advanced Boiling Water Reactor,''
b. Appendix D to 10 CFR part 52, ``Design Certification Rule for
the AP1000 Design,''
c. Appendix E to 10 CFR part 52, ``Design Certification Rule for
the ESBWR Design,''
d. Appendix F to 10 CFR part 52, ``Design Certification Rule for
the APR1400 Design,'' and
e. Appendix G to 10 CFR part 52, ``Design Certification Rule for
NuScale.''
Two design certifications have expired because no timely renewal
applications were submitted in accordance with Sec. 52.57,
``Application for renewal.'' These are appendix B to 10 CFR part 52,
``Design Certification Rule for the System 80 + Design,'' and appendix
C to 10 CFR part 52, ``Design Certification Rule for the AP600
Design.'' Accordingly, this proposed rule would apply to the five
design certifications currently in effect and would not apply to the
two expired design certifications.

Parts 52 and 53--Notification of Initial Fuel Load

The current regulations in Sec. Sec. 52.103(a) and 53.1452(a)
require an applicable licensee to notify the NRC of the licensee's
scheduled date for initial loading of fuel at least 270 days before the
scheduled date and subsequently to notify the NRC of any updates to the
schedule every 30 days thereafter. In those cases where, after
providing the initial notification, the licensee experiences extended
delays, the licensee would still be required to provide updates every
30 days even if the schedule no longer changes. The regulations in
Sec. 53.1452(a) also require that COL holders installing fueled
manufactured reactors notify the NRC at

[[Page 60714]]

least 270 days prior to initiating the removal of the features to
prevent criticality and subsequently to notify the NRC of any updates
to the schedule every 30 days thereafter. This proposed rule would
eliminate the requirement for recurrent reporting of updates to the
licensee's scheduled date for initial fuel load every 30 days after the
initial report submitted no less than 270 days before the scheduled
date. Instead, this would be replaced with a requirement to notify the
NRC only in those cases where, after providing the initial
notification, the scheduled date for initial fuel load is revised.

10 CFR Part 52--Requirements for Evaluation of Generic Issues and
Operating Experience

The regulations in Sec. Sec. 52.47(a)(21) and (22), 52.79(a)(20)
and (37), 52.137(a)(21) and (22), and 52.157(f)(28) and (29) require
applicants to submit, for NRC review, certain generic issues and
operating experience from generic communications technically relevant
to the associated design. As explained in Regulatory Guide 1.206,
``Applications for Nuclear Power Plants,'' Revision 1, October 2018,
the scope of the generic issues to be evaluated is those Unresolved
Safety Issues (USIs) and medium- and high-priority Generic Safety
Issues (GSIs) that (1) are identified in the version of NUREG-0933, ``A
Prioritization of Generic Safety Issues,'' current on the date up to 6
months before the docket date of the application and (2) are
technically relevant to the design. RG 1.206, Revision 0, ``Combined
License Applications for Nuclear Power Plants,'' also states that the
scope of the operating experience from generic communications to be
evaluated is those Generic Letters and Bulletins issued after the most
recent revision of NUREG-0800, ``Standard Review Plan for the Review of
Safety Analysis Reports for Nuclear Power Plants: LWR Edition,'' and 6
months before the docket date of the application, or comparable
international operating experience.
The Commission's Policy Statement on Severe Reactor Accidents
Regarding Future Designs and Existing Plants (50 FR 32138; August 8,
1985) and Policy Statement on Standardization (52 FR 34884; September
15, 1987) introduced the policy of evaluating USIs and medium- and
high-priority GSIs, including a special focus on assuring the
reliability of decay heat removal systems and the reliability of
electrical supply systems. The 1989 10 CFR part 52 rulemaking (54 FR
15372; April 18, 1989) codified the policy, and the subsequent 10 CFR
part 52 rulemaking (72 FR 49352; August 28, 2007) introduced the
provisions related to operating experience.
As discussed in SECY-15-0002, ``Proposed Updates of Licensing
Policies, Rules, and Guidance for Future New Reactor Applications''
(ML13277A420), while 10 CFR part 50 does not contain regulations
identical to those in 10 CFR part 52, under Commission policy,
applicants under 10 CFR part 50 demonstrate the resolution of USIs and
GSIs prior to issuance of an operating license.
The NRC proposes to remove this codified requirement for 10 CFR
part 52 applicants, and to ensure consistency across frameworks,
applicants under 10 CFR part 50 would not be required to demonstrate
the resolution of USIs and GSIs prior to issuance of an operating
license as well. These applicants would still be required to
demonstrate that applicable NRC safety requirements are satisfied. Part
53 does not include requirements related to the evaluation of the
generic issues and operating experience as does 10 CFR part 52.
The evaluation of the generic issues and operating experience from
generic communications was intended to lead to improved efficiency and
effectiveness of the NRC safety reviews. However, experience with new
reactor licensing has demonstrated that these requirements have led to
applicants expending significant resources without significant
contributions to either the effectiveness or efficiency of the
associated safety review. Accordingly, the evaluation is an unnecessary
burden given the limited benefit. This assessment result is largely
based on the following.
First, as the subject generic issues and generic communications
were resolved, regulations and guidance have typically been revised or
developed to inform future safety reviews. The regulatory guidance,
e.g., the Standard Review Plan and Advanced Reactor Content of
Application Project (ARCAP) guidance (<a href="https://www.nrc.gov/reactors/new-reactors/advanced/modernizing/guidance/advanced-reactor-content-of-application-project.html">https://www.nrc.gov/reactors/new-reactors/advanced/modernizing/guidance/advanced-reactor-content-of-application-project.html</a>), regulatory guides, and endorsed industry
consensus standards, contain provisions and acceptance criteria that
reflect the lessons learned from past operating experience including
generic issues and generic communications. In addition, as the NRC's
approach to new and advanced reactor licensing has substantially
evolved, the consideration of operating experience, generally broader
than those from generic issues and generic communications, are further
integrated into both NRC and industry guidance and practices. An
example is the advancement and increased use of probabilistic risk
assessment (PRA), which is required under 10 CFR part 52. The industry
PRA standard, endorsed by the NRC in regulatory guides, involves an
evaluation of operating experience for various technical elements
including initiating event analysis and event sequence analysis with an
emphasis on a systematic and comprehensive assessment of hazards and
associated licensing events of potential risk or safety significance.
Accordingly, licensing experience has demonstrated that a separate
evaluation of the subject generic issues and generic communications was
not necessary as it largely became redundant.
Next, most past generic issues and generic communications are based
on the operating LWRs. Their applicability and technical relevance to
new and advanced reactors, such as small modular reactors and non-LWR
reactors, are very limited. These new and advanced reactor designs,
often much simpler and safer using inherent features and passive
systems, have eliminated or reduced the potential risk that,
historically, was addressed via evaluation of operating experience,
including those identified in the generic issues or generic
communications. The introduction of advanced reactor designs, such as
microreactors, is expected to further limit the applicability and
technical relevance of the subject generic issues and generic
communications. Further, new and advanced reactor designs are
significantly different from the operating reactors and reflect the
vast operating experience from the design and operation of nuclear
reactors of earlier generations. For example, new and advanced reactors
provide enhanced margins of safety and/or use simplified, inherent,
passive, or other innovative means to accomplish their safety
functions, consistent with the Commission's Policy Statement on the
Regulation of Advanced Reactors (73 FR 60612; October 14, 2008).
Finally, since its inception in 1976, the NRC's Generic Issues
Program addresses potential safety or security concerns that affect
multiple nuclear facilities and are not adequately covered by existing
regulations or programs. Most of the generic issues have been
identified during the plant operations. USIs are a specific subset of
generic issues that are considered to have potential safety
significance and require further analysis before a regulatory decision
can be made. Over the years, fewer generic issues have been

[[Page 60715]]

identified, with the most recent issue (GSI-206: Flooding of Nuclear
Power Plant Sites Following Upstream Dam Failures) being identified in
2011. Currently, all generic issues have been resolved, generally
meaning that regulations or programs are in place for future licensing
reviews. Similarly for generic communications, the issuance of generic
letters and bulletins has decreased over the years, with the most
recently issued Bulletin being in 2012 and Generic Letter in 2016. With
sections of NUREG-0800 being updated occasionally, those Bulletins and
Generic Letters issued since the last update of NUREG-0800 are minimal
or none, especially those that may be technically relevant to new and
advanced reactors. The underlying issues associated with Generic
Letters and Bulletins, which often involves an NRC information request
from the affected licensees, are typically resolved as part of the
process implementation of NRC MD 8.18, ``Generic Communication
Program.''
Accordingly, this rulemaking would eliminate the regulations under
10 CFR part 52 that require applicants to submit for NRC review an
evaluation of generic issues and operating experience from generic
communications technically relevant to their design.

10 CFR Parts 50, 52, and 53--Standard Design Approvals

Under current regulations, Sec. Sec. 52.147 and 53.1218, SDAs
expire 15 years after issuance. This expiration period is inconsistent
with the duration of other approvals, such as design certifications and
combined licenses, that are also supported by renewal provisions.
Moreover, SDAs are currently excluded from the change control
processes outlined in Sec. Sec. 50.59, ``Changes, tests, and
experiments,'' and 53.1550, ``Evaluating changes to facility as
described in Final Safety Analysis Reports.'' They also lack a
mechanism for amendments requiring prior NRC approval, limiting the
ability of SDA holders to update their designs without initiating a new
approval process.
Accordingly, this rulemaking addresses stakeholder feedback and
supports the NRC's strategic objectives to modernize its regulatory
framework, reduce unnecessary regulatory burden, and foster innovation
in reactor design by amending 10 CFR 52.147 and 53.1218 to remove the
expiration date for SDAs.

10 CFR Parts 52 and 53--Environmental Reviews of Early Site Permits

At a high level, the National Environmental Policy Act of 1969, as
amended (NEPA), requires Federal agencies, including the NRC, to
consider significant effects of a major Federal action on the quality
of the human environment. In general, this process involves determining
the appropriate level of review (i.e., whether an environmental
document such as an environmental impact statement or an environmental
assessment is required or whether a categorical exclusion applies). As
relevant to this rulemaking, the NRC's current regulations in
Sec. Sec. 52.18 and 53.1149(a), both titled, ``Standards for review of
application,'' require that the NRC prepare an environmental impact
statement as part of its review of an application for an early site
permit, regardless of whether a lesser level of review would be more
appropriate. As such, this proposed rule would no longer require the
NRC to always prepare an environmental impact statement for early site
permits. Instead, the NRC would have the flexibility to assess the
appropriate level of its environmental reviews on a case-by-case basis,
commensurate with the reasonably foreseeable effects on the quality of
the human environment. This proposed change would result in internal
efficiencies and reduce resource burdens, including lesser burdens on
applicants to provide supporting information, where preparation of an
environmental impact statement would not be warranted.

10 CFR Part 50--Post-TMI Requirements

As discussed in SECY-22-0052 and as approved in SRM-SECY-22-0052,
the NRC is proposing to amend paragraph (f), ``Additional TMI-related
requirements'' of Sec. 50.34, ``Contents of applications; technical
information.'' Specifically, Sec. 50.34(f) would be amended to require
10 CFR part 50 applicants to provide information related to addressing
lessons learned from the Three Mile Island (TMI) accident in the same
manner as currently required for 10 CFR part 52 applicants. These
changes would improve clarity or remove unnecessary requirements and
ensure consistency in new reactor licensing reviews across licensing
frameworks.

10 CFR Parts 50, 53, and 55--Updates to the Operator Licensing Programs

A. Minimum Number of Manipulations
In 10 CFR part 55, ``Operators' Licenses,'' Sec. 55.31(a)(5)
currently requires that operator license applicants, as trainees,
provide evidence of successfully manipulating the controls of either
the plant for which the license is sought or a simulator. It further
specifies that at least five significant control manipulations must be
performed that affect reactivity or power level.
The NRC identified Sec. 55.31(a)(5) as a regulation that should be
considered for revision because it is prescriptive, based on currently
operating reactors, and may not be needed for safety in all cases
(e.g., for new advanced reactor designs). The NRC is proposing to
change Sec. 55.31(a)(5) to add a provision that in lieu of five
significant control manipulations, the facility licensee may determine
the minimum number of required control manipulations in accordance with
its Commission-approved training program. This change would improve
flexibility in those situations and ensure that an appropriate
alternative is established.
B. Alternative Examination Standards
The NRC would revise paragraph (a) and delete paragraph (d) of
Sec. 55.40, ``Implementation,'' to allow the use of criteria other
than the criteria in NUREG-1021, ``Operator Licensing Examination
Standards for Power Reactors,'' and NUREG-1478, ``Operator Licensing
Examiner Standards for Research and Test Reactors,'' to avoid the NRC
having to revise the NUREGs for each new plant design and/or licensees
having to submit exemptions that seek to use a document other than
these NUREGs. Requiring the use of these NUREGs has been the means of
complying with the AEA, which requires the Commission to ``prescribe
uniform conditions for licensing individuals as operators'' for
different classes of facilities, but the statutory requirement can
still be achieved without having the examination standards for every
plant design contained in a NUREG. For example, facility licensees
could submit criteria for NRC review in an examination standards
topical report.
C. Written Examinations and Operating Tests
The NRC would revise Sec. Sec. 55.41(b), 55.43(b), and 55.45(a) to
allow greater flexibility and clarity for the content of written
examinations and operating tests. Currently, these regulations require
that written examinations and operating tests include a representative
sample of the topics listed therein, to the extent applicable to the
design of the facility. These topics may not include all of the safety-
significant topics applicable to plant designs that are not water-
cooled reactors, so there may be other safety-significant topics that
should be sampled on written examinations and operating tests for

[[Page 60716]]

operator license applicants at non-LWR plants but are not currently
covered by these regulations. This could create ambiguity and
confusion, resulting in exemption requests. The proposed edits to
Sec. Sec. 55.41(b), 55.43(b), and 55.45(a) would address these
concerns.
D. Simplification of Simulator Requirements
The NRC would remove the terms ``plant-referenced simulator'' and
``Commission-approved simulator'' from 10 CFR part 55. The term
``simulation facility'' would remain, and its definition in Sec. 55.4,
``Definitions,'' would be revised. The NRC would not add any new
requirements for simulation facilities; the performance-based
requirements that a simulation facility must meet in order to be used
for operating tests, requalification, or meeting experience
requirements would remain in Sec. 55.46, ``Simulation facilities.''
These changes would clarify for facility licensees of reactors under
construction the requirements that a simulation facility must meet to
be used for operating tests, requalification, or meeting experience
requirements. As long as a simulation facility meets these proposed
requirements, it could be used for those purposes; there would be no
need to determine whether the simulation facility is a plant-referenced
simulator or if it instead must be approved as a simulation facility by
the Commission. The NRC would confirm that a simulation facility
complies with regulatory requirements by inspecting simulator
performance test results and other simulator data as necessary.
Additionally, the NRC has observed an increased use by the industry of
``glasstop'' simulators in training. ``Glasstop'' simulators consist of
one or more touch-sensitive panels that mimic the analog and digital
controls, alarms, and displays found in the main control room of a
nuclear power facility. They do not look like a full-scale control
room, but they can replicate plant systems accurately. The proposed
changes would clarify for facility licensees of operating reactors the
requirements to use glasstop simulators for additional purposes, such
as conducting portions of the operating tests required by Sec. 55.45.
As long as a glasstop simulator meets the proposed requirements, it
could be used for those purposes; there would be no need to determine
whether the glasstop simulator is a plant-referenced simulator or if it
instead must be approved as a simulation facility by the Commission.
The NRC would confirm that a simulation facility complies with
regulatory requirements by inspecting simulator performance test
results and other simulator data as necessary. Minor conforming changes
would be made to Sec. 53.725(c).
E. Requalification Examination and Program Periodicity Requirements
In Sec. 55.59(a), ``Requalification requirements,'', the NRC
requires each licensed operator to successfully complete a
requalification program developed by the facility licensee and to pass
a comprehensive written examination and an annual operating test as
part of the requalification program. Paragraph (h) of Sec. 55.53,
``Conditions of license,'' also contains a condition applicable to all
operators' licenses that the licensee shall complete the
requalification program, which includes passing the annual operating
tests and biennial written examination. This is redundant to Sec.
55.59(a).
Section 55.59(a) also lists the topics that must be included on the
written examination and operating test; however, this requirement is a
program requirement, and it would be more appropriate to include under
the program evaluation requirements in Sec. 55.59(c)(4),
``Evaluation.''
Additionally, in Sec. 55.59(c), ``Requalification program
requirements,'' the NRC requires that the requalification program be
conducted for a continuous period not to exceed 2 years, and that each
2-year program (i.e., cycle) be followed by a subsequent 2-year
program. The NRC proposes to allow a facility licensee, using the
systems approach to training process, to appropriately determine a
different length of time for its requalification program and different
requalification examination periodicities.
The NRC identified the foregoing regulations as ones that should be
considered for revision because they are prescriptive and may not be
needed for safety in all cases. The proposed changes to Sec. 55.59
would improve flexibility by allowing alternatives where they are
justified by the facility licensee in its approved requalification
program.
Conforming changes are proposed to 10 CFR part 53. Furthermore, the
NRC proposes to amend Sec. 53.730(g)(3) to align requalification
program implementation requirements to those in Sec. 50.54(i-1).
Currently, Sec. 53.730(g) states that requalification programs must be
approved by the NRC as part of its approval of the OL or COL for the
plant, and Sec. 53.730(g)(3) requires that the requalification program
be implemented upon commencing the administration of initial
examinations. However, the administration of initial examinations could
potentially begin before an OL is required to be approved. The proposed
change would prevent this situation.
F. Requirements for Licensed Medical Examiners
The NRC would revise Sec. Sec. 55.21, ``Medical examination,'' and
55.23, ``Certification,'' to allow either a physician or a licensed
medical examiner to perform and certify the results of the medical
examinations required for licensed operators and applicants for
operator licenses. Currently, the medical examinations may only be
certified by a physician, as defined in Sec. 55.4, meaning an
individual licensed to dispense drugs in the practice of medicine.
The NRC identified these requirements as anticompetitive and overly
restrictive. The NRC proposes to allow a licensed medical examiner
(i.e., a medical professional, other than a physician, licensed by
their state to independently perform medical examinations for the
purpose of determining an individual's medical health and general
condition) to perform and certify the results of the medical
examinations required for licensed operators and applicants for
operator licenses. This proposed change would reduce burden by
expanding the pool of available medical professionals, especially in
rural areas where doctors already may be in relatively shorter supply.
If more nuclear power plants are constructed, more operators will be
needed, and they will need medical examinations. Allowing medical
professionals licensed by their state to perform medical examinations
to do the operator examinations would likely help alleviate demand for
doctors, especially in areas where they may be in short supply.
Conforming changes would be implemented in Sec. Sec. 53.765 and
53.775(b)(2).
G. Changes to Treatment of Medical License Conditions
The NRC would revise Sec. Sec. 55.25, ``Incapacitation because of
disability or illness,'' 55.27, ``Documentation,'' 55.33, ``Disposition
of an initial application,'' and 55.53 to allow for certain medical
conditions to be covered under a new, generic condition of licenses in
Sec. 55.53. Currently, applicants for a license must have a medical
examination that must be certified by a physician on NRC Form 396,
``Certification of Medical Examination by Facility Licensee.'' The
certifying physician must determine that the applicant either meets all
health

[[Page 60717]]

requirements or, if the applicant does not meet all health
requirements, that any medical conditions can be accommodated through
specific license conditions (e.g., a license condition that says
something akin to, ``you shall wear corrective lenses while performing
licensed operator duties,'' for applicants who do not meet the minimum
vision requirements). If the applicant's Form 396 indicates that the
applicant has no medical issues requiring accommodation via a license
condition, then the applicant can be found by the Commission to meet
the requirements of Sec. 55.33(a)(1). If the applicant's Form 396
indicates that the applicant has one or more issues that can be
accommodated by one or more license conditions, then the Commission may
issue a license with the specific license conditions that are necessary
to accommodate the medical condition(s) (e.g., ``you shall take
medications as prescribed,'' for an applicant who has diabetes that is
controlled by medication). Once a license is issued, the license must
be amended if either any new, permanent medical conditions arise that
need to be accommodated by one or more license conditions, or if a
medical condition changes such that one or more license conditions are
no longer necessary (e.g., the operator gets LASIK surgery and no
longer needs corrective lenses, rendering a license condition to wear
corrective lenses unnecessary).
The NRC uses a qualified medical expert to review certain medical
documents for applicants and operators to ensure that the license
conditions proposed by the examining physician are sufficient to
accommodate the identified medical conditions. The process is as
follows:
a. For new applications, the NRC reviews the applicant's NRC Form
396.
b. If corrective lenses, hearing aids, and/or use of a continuous
positive airway pressure device are checked as necessary license
conditions, then the NRC will not send the documentation forward for
review by a qualified medical expert. The NRC will then issue a license
with the corresponding specific license condition(s).
c. If any other license conditions are identified as being
necessary, then the NRC sends the documentation to a qualified medical
expert, who reviews the applicant's medical information (i.e., the
``medical evidence'' submitted on NRC Form 396) to make sure that the
proposed license condition(s) will accommodate the medical
condition(s). The qualified medical expert documents the results of his
or her review and provides it to the NRC. The NRC issues the license
with the necessary, specific conditions listed on the license.
d. If a licensed operator develops a new, permanent medical
condition that must be accommodated by one or more license conditions,
then they must inform the NRC and submit a new NRC Form 396 to request
any new, specific license conditions be added to the license. The NRC
uses the same process described above for new applications for license
amendments.
To streamline this process, the NRC proposes to add a new, generic
condition of licenses to Sec. 55.53 that would eliminate the need for
specific license conditions for certain medical restrictions (i.e.,
corrective lenses, hearing aids, and use of a continuous positive
airway pressure machine). The proposed new generic condition in Sec.
55.53 would state that a licensee must comply with all license
conditions identified on NRC Form 396 that are designated as covered by
Sec. 55.53(m), and NRC Form 396 would identify the medical
restrictions (i.e., license conditions) that are covered by Sec.
55.53(m). The number of licenses with specific conditions for
corrective lenses and/or hearing aids that could be replaced with the
proposed new generic license condition are as follows per region/office
(note: license conditions to use a therapeutic device are not included
in the count below since they may or may not be limited to continuous
positive airway pressure machines, so these numbers are slightly lower
than the numbers of specific license conditions that would actually be
able to be replaced):

<bullet> Headquarters: 146 (71% of all licenses with specific medical
conditions)
<bullet> Region I: 192 (40% of all licenses with specific medical
conditions)
<bullet> Region II: 207 (25% of all licenses with specific medical
conditions)
<bullet> Region III: 159 (30% of all licenses with specific medical
conditions)
<bullet> Region IV: 148 (30% of all licenses with specific medical
conditions)

On average, about 39% of licenses with specific license conditions
would be ``covered'' under the proposed new general license condition
in Sec. 55.53(m).
Furthermore, there has been ambiguity for many years regarding what
is considered a ``permanent'' medical condition. The NRC receives
correspondence from facility licensees regarding temporary medical
conditions; these correspondence are not required. The NRC proposes to
amend Sec. Sec. 55.25 and 55.27 to specify how facility licensees
should accommodate temporary medical conditions, as distinct from
permanent medical conditions, and to clarify that these do not need to
be treated in the same manner as permanent medical conditions.
H. Changes to Operator License Reporting Requirements
As discussed in Sec. 50.74, ``Notification of change in operator
or senior operator status,'' three reports are currently required for
changes to operator status. NEI discussed these reports in a May 6,
2020, letter on the Retrospective Review of Administrative Requirements
(RROAR) (ML20128J340) and recommended deleting them. The reporting
requirements in Sec. 50.74(a) and (b) address the situations where the
10 CFR part 55 license holder's employment at the facility has been
terminated or is no longer needed. The NRC does not propose to remove
these two reporting requirements. The facility licensee should be
responsible for reporting the required information to the NRC because
(1) under this proposed rule, operator licenses would no longer be
issued with license terms, and (2) the NRC needs to know the status of
the operator licenses that it issues (including who actually holds
operator licenses and, thus, is required to meet the requalification
requirements and medical examination requirements). However, the NRC
proposes to improve clarity by including these reporting requirements
in 10 CFR part 55 instead of in 10 CFR part 50. The third reporting
requirement, in Sec. 50.74(c), is proposed to be deleted because it is
redundant to Sec. Sec. 55.25, and 55.25 would be amended to remove
reference to Sec. 50.74 and replace it with Sec. 55.5,
``Communications.''
Additionally, Sec. 55.53(g) requires a licensed operator to notify
the Commission within 30 days of conviction of a felony, and Sec.
55.59(b) requires a licensed operator to submit evidence of completion
of additional training to the Commission when necessary. The NEI letter
dated May 6, 2020, regarding RROAR included a proposed change to Sec.
55.53(g) to remove the burden of notifying the NRC from the individual
operator. The NRC agrees and proposes to amend these sections to allow
licensed operators to provide information to the NRC via the facility
licensee rather than requiring the operator to provide it to the NRC
directly. Conforming changes for Sec. 55.53(g) would be made to
Sec. Sec. 53.785(h) and 53.810(f). Conforming changes for Sec.
55.59(b) are not needed in 10 CFR part 53.

[[Page 60718]]

I. Revision to Training Requirements in Sec. 50.120
Paragraph (b)(1) in Sec. 50.120, ``Training and qualification of
nuclear power plant personnel,'' incudes an 18-month timeframe, based
on the date of fuel load, for the establishment, implementation, and
maintenance of a training program that meets the requirements of Sec.
50.120(b)(2) and (3). This timeframe is not necessary for safety, and
since the requirement was first established, ambiguity has existed as
to what exactly is required to meet the 18 months before fuel load
criteria. Therefore, the NRC proposes to change Sec. 50.120 to clarify
that the requirement to establish, implement, and maintain the training
program must be met with sufficient time to ensure that qualified
personnel are available to do their assigned jobs. This proposal was
also included in a letter from NEI titled, ``Industry Recommendations
on Accelerating NRC Reform,'' dated July 31, 2025 (ML25212A197).
Additionally, Sec. 50.120 prescribes that specific training
programs be implemented, including for a shift technical advisor. In
2020, NuScale submitted a topical report, which was approved by the NRC
on May 26, 2021 (ML21231A286), to support future exemptions from the
requirement to have a shift technical advisor training program because
NuScale plants would not have a shift technical advisor. The NRC
proposes to add ``as applicable'' to the list of personnel to be
covered by the training program so as to avoid the need for future
exemptions in cases where certain categories of plant personnel are
determined not to be necessary at a plant.
Section 53.830, ``Training and qualification of commercial nuclear
personnel,'' is already similar to proposed Sec. 50.120, with a minor
difference: Sec. 53.830 states that prior to initial fuel load (or,
for a fueled manufactured reactor, prior to initiating the removal of
the features to prevent criticality required under Sec. 53.620(d)(1)),
each holder of an OL or COL must, with sufficient time to provide
trained and qualified personnel to operate the facility, establish,
implement, and maintain a training program that demonstrates compliance
with the requirements of Sec. 53.830(c) and (d). It imposes two time-
based requirements to have the training programs established: prior to
fuel load/initiating removal of features to prevent criticality, and
with sufficient time to provide qualified personnel to operate the
facility. The NRC proposes to remove the former requirement because
there is no reason to also require the training program to be
implemented prior to fuel load, as this activity would require
sufficiently trained and qualified or even licensed plant personnel,
and that would be covered by the requirement to have trained and
qualified personnel in time to perform jobs that require trained and
qualified personnel.
J. Elimination of Wait Times for Re-Applications in Sec. 55.35
Paragraph (a) of Sec. 55.35, ``Re-applications,'' requires an
applicant for an operator license whose application has been denied
because of failure to pass the written examination, operating test, or
both, to wait 2 months after the first denial, 6 months after a second
denial, and then 2 years for each subsequent denial. These wait times
are unnecessary and create additional burden for operator license
applicants and facility licensees. Therefore, the NRC proposes to amend
this requirement to allow applicants to file re-applications regardless
of the time that has elapsed since the previous application denial.
K. Operating Tests
In Sec. 55.40, the NRC currently provides the option for power
reactor facility licensees to prepare, proctor, and grade the written
examinations required by Sec. Sec. 55.41, ``Written examination:
Operators,'' and 55.43, ``Written examination: Senior operators,'' and
to prepare the operating tests required by Sec. 55.45, ``Operating
tests,'' subject to conditions specified therein. Section 55.40 also
says that the NRC shall prepare, proctor, and grade examinations and
tests for non-power reactor facility licensees. These regulations
became effective on October 20, 1999, in the final rule, ``Initial
Licensed Operator Examination'' (64 FR 19868). The Commission did not
then permit licensees to administer and grade operating tests (with NRC
examiners observing facility employees) in part due to a concern with
the reduced NRC involvement and the potential for impact on
independence, objectivity, and public perception. In SECY-96-206,
``Rulemaking Plan for Amendments to 10 CFR part 55 to Change Licensed
Operator Examination Requirements'' (ML20134C900), the NRC staff
discussed the option of having NRC examiners observe facility employees
conduct the simulator scenarios and a sample of the walk-through tasks
for the operating tests for each applicant: ``[This option] would have
had the least NRC involvement and, therefore, the greatest potential
impact on independence and public perception. By not directly observing
each applicant, the NRC examiner would be in a weaker position to judge
the performance of the applicant so as to determine which applicants
should pass or fail if a crew responded incorrectly during a simulator
scenario.'' In SRM-SECY-96-206, dated December 17, 1996 (ML20132C429),
the Commission directed the NRC staff to continue to administer and
grade the operating tests.
Today, nearly all power reactor operators' license written
examinations and operating tests are prepared by facility licensee
staff, and power reactor operators' license written examinations are
frequently proctored and graded by facility licensee staff. As required
by current regulations, operating tests are administered by an NRC
examination team with the support of facility licensee personnel (e.g.,
facility licensee personnel run the simulator used for power reactor
simulator scenarios, taking direction from the NRC chief examiner on
when the simulator should initiate the events per the pre-approved
simulator examination guide), and the NRC examiners conduct the grading
of their designated applicants, with no input from facility licensee
personnel.
Certain power and non-power reactor stakeholders have recently
indicated an interest in also proctoring and grading the operating
tests for their operator license applicants, such that the NRC's role
would shift from administering the operating tests and grading
applicants to overseeing the facility licensee staff in its
administration and grading of the operating tests. In addition, in its
July 31, 2025, letter to the NRC, NEI recommended this change.
Moreover, industry now possess substantial experience in
administering, proctoring, and grading examinations as part of both the
Commission-approved INPO-accredited training process--which includes
audit examinations that closely model NRC initial exams--and the long-
standing NRC-mandated requalification program, which requires licensees
to administer and grade annual operating tests and biennial written
examinations subject to NRC inspection and oversight. Although the
industry has not prepared the operating test grading documentation for
the NRC initial operator licensing examination, the staff has developed
detailed guidance in NUREG-1021 to support licensee preparation and
documentation of initial exam grading, and NRC examiners would continue
to oversee and evaluate licensee performance, enabling prompt
identification and correction of any gaps that may arise during
implementation.

[[Page 60719]]

Section 55.40 is currently silent regarding who will administer and
grade the operating tests for power reactor facilities, but it is
implied that only the NRC will do that given that the regulation
specifies other activities that power reactor facilities may do (and it
is explicitly stated that way in NUREG-1021, revision 12). Therefore,
this proposed rule would amend Sec. 55.40 to specifically allow
facility licensees to administer and grade operating tests subject to
certain conditions.
L. Plant Walkthrough Requirements in Sec. 55.45
In Sec. 55.45(b), ``Implementation--Administration,'' the NRC
requires that the operating test be administered in a plant walkthrough
and in either a simulation facility that the Commission has approved
for use after application has been made by the facility licensee under
Sec. 55.46(b), ``Commission-approved simulation facilities and
Commission approval of use of the plant in the administration of the
operating test,'' a plant-reference simulator, or the plant, if
approved for use in the administration of the operating test by the
Commission. Section 55.40(a) requires, in part, that the NRC use the
criteria in NUREG-1021 to prepare the operating tests required by Sec.
55.45, and to evaluate the operating tests prepared by power reactor
facility licensees. As discussed in NUREG-1021, revision 12, section
ES-3.1, ``Overview of the Operating Test for Operator Licensing Initial
Examinations,'' the walkthrough portion of the operating test consists
of a set of job performance measures. Each applicant for an operator's
license must complete a certain number of job performance measures in
the plant.
The plant walkthrough requirement was the subject of exemption
requests for Vogtle Electric Generation Plant (VEGP) Unit 3 and for
Virgil C. Summer Nuclear Station (VCSNS) Unit 2. In the June 30, 2016,
Federal Register notice (FRN) of the NRC's granting approval of the
exemption for VEGP Unit 3 (81 FR 42745) and in the August 22, 2016, FRN
for the NRC's granting approval of the exemption for VCSNS Unit 2 (81
FR 56704), the NRC also approved alternatives to the in-plant methods
of testing described in NUREG-1021. Specifically, the NRC approved the
use of discussion and performance evaluation methods in combination
with plant layout diagrams, maps, equipment diagrams, pictures, and
mock-ups while the plant is under construction.
A substantial number of a plant's systems must be constructed
before doing a plant walkthrough during the NRC initial licensing
examination so that the plant walkthrough portion of the operating test
is not predictable. Predictable examinations may prevent the examiner
from distinguishing applicants who have mastered the required knowledge
and skills from those who have not. However, the completion of plant
construction occurs relatively close in time to the scheduled date for
fuel loading and later operation of the facility. Thus, administration
of the NRC initial licensing examinations would need to occur
relatively close in time to the scheduled date for fuel loading at the
plant. Delaying the administration of the NRC operating test, which
includes in-plant job performance measures, until plant construction is
complete is not desirable.
The NRC examination will likely overlap with preoperational testing
activities; applicants may be required to participate in preoperational
testing at the same time the NRC examination is administered.
Insufficient NRC examination throughput (i.e., the number of applicants
that pass the NRC examination) could cause preventable delays in the
facility licensee's ability to begin fuel loading. There might be
missed opportunities for early identification and timely incorporation
of lessons learned into the operator training and licensing process
without a safe and deliberate approach to administering examinations to
smaller groups of applicants in succession. The NRC's ability to
administer examinations for the large number of applicants necessary to
staff a new single large LWR unit (40-50 operators) in time for fuel
load, while also carrying out examinations and inspections at operating
reactor sites, could be challenged.
Therefore, the NRC proposes to amend the plant walkthrough
requirements in Sec. 55.45(b) to allow alternatives to the in-plant
methods of testing while the plant is under construction.
M. Examination Waiver Requirements
Under Sec. 55.47, ``Waiver of examination and test requirements,''
the NRC may waive any or all of the requirements for a written
examination and operating test. As discussed in SECY-22-0052 and
approved in SRM-SECY-22-0052, the NRC is proposing changes to the
waiver criteria to allow waivers in the circumstances of adding
additional units to a standard design. Additionally, Sec. 53.780(f)
would be revised with conforming changes.
N. Removing Obsolete Information From Sec. 55.5
In Sec. 55.5, instructions are provided for submitting information
to the NRC. Currently, the electronic information exchange (EIE) is
listed as an example of an electronic submission method. However, the
NRC plans to transition away from the EIE and replace it with other
options for the electronic submission of documents (e.g., the NRC
external portal). Therefore, the NRC proposes to remove the term
``EIE'' throughout 10 CFR part 55 as an example of a method of
electronic submission.
O. Elimination of License Renewal for 10 CFR Part 55 Operators'
Licenses
Historically, the requirement for renewal of operators' licenses
afforded the NRC the opportunity to re-evaluate the licensing basis of
licensed operators. The purpose of license renewal was to assess the
likelihood of an operator's continued ability to safely operate the
facility.
The NRC identified that eliminating license terms and, thus, the
requirement for license renewal, would meet goals stated in E.O. 14300
and E.O. 14192, ``Unleashing Prosperity Through Deregulation.''
Specifically, if the number of nuclear plants and licensed operators
were to increase, the amount of renewals would also increase, requiring
an expenditure of more licensee and NRC resources. Removing the
requirement for licensed operators to request renewal of their licenses
every 6 years would be a reduction in burden for the licensed
operators, facility licensees, and the NRC. Therefore, the NRC is
proposing to eliminate license terms and the requirement for license
renewal. Conforming changes would also be made to 10 CFR part 53.
The NRC has previously reconsidered license terms: in December
2024, the NRC eliminated license terms for certain NPUFs (see 89 FR
106234). The NRC determined that the overall objectives of license
renewal can likely be achieved through existing oversight activities
while eliminating the administrative burden associated with renewal.
This proposed approach would be consistent with the NRC's goal of
efficient and effective licensing. Additionally, this is not the first
time that the NRC has amended renewal requirements for operators'
licenses: license terms were extended from 2 years to 6 years in the
final rule, ``Operators' Licenses and Conforming Amendments,'' issued
on March 25, 1987 (52 FR 9453).

[[Page 60720]]

P. Generally Licensed Reactor Operators for Self-Reliant Mitigation
Facilities
Part 53, subpart F, describes a class of facility, the self-reliant
mitigation facility (SRMF). For a facility to be classified as a SRMF,
the NRC must find that certain criteria established in 10 CFR part 53
have been met. For SRMFs, specific licenses for operators (i.e.,
individual operator and senior operator licenses) are not required.
Instead, the facility licensee may rely on a general license for
reactor operators issued by the NRC under 10 CFR part 53.
Part 53 is an optional framework, and 10 CFR parts 50 and 52 are
also options for an NRC applicant. As of December 2025, three NRC
applicants that are using or intend to use the current part 50
licensing process had proposed that their facilities are SRMFs, and,
accordingly, expressed an interest in using the same provisions for a
general license for reactor operators that are in 10 CFR part 53.
Currently, there are no provisions for generally licensed reactor
operators (GLROs) in the NRC's regulations other than in 10 CFR part
53. Accordingly, any applicant for an SRMF would need to request
exemptions to use an approach comparable to a general license for
reactor operators if it is not using 10 CFR part 53. Such an approach
would be less efficient than adding to 10 CFR part 55 provisions
equivalent to those in 10 CFR part 53. Therefore, the NRC proposes to
add to 10 CFR part 55 optional provisions for GLROs, which would be
applicable to 10 CFR parts 50 and 52 facility licensees of SRMFs.
Adding provisions for GLROs to 10 CFR part 55 would also require
revising 10 CFR parts 50 and 52 to establish associated requirements
related to staffing for GLROs akin to those requirements in 10 CFR part
53. The licensed operator staffing requirements in Sec. 50.54(m) that
apply to OL or COL holders do not account for GLROs and are not always
necessary for safety.
Minor editorial conforming changes would be made to 10 CFR part 53.

10 CFR Part 75--IAEA Notifications

Paragraphs (b)(3) and (c) of Sec. 75.13, ``Safeguards on Nuclear
Material--Implementation of Safeguards Agreements Between the United
States and the International Atomic Energy Agency,'' provide that a
``request'' for non-disclosure of information to the public made under
Sec. 2.390, ``Public inspections, exemptions, requests for
withholding,'' not be treated as a ``request'' made under Sec. 75.13
for confidential information not to be transmitted physically to the
IAEA.
This rulemaking proposes eliminating Sec. 75.13(c) because it is
redundant with Sec. 75.13(b)(3), as both contain identical language
except that Sec. 75.13(b)(3) references Sec. 2.390 while Sec.
75.13(c) references Sec. 2.390(b).

10 CFR Parts 50, 52, 53, 70, and 72--Record Keeping and Reporting
Requirements

On February 4, 2020, the NRC published a 60-day notice in the
Federal Register (85 FR 6103) requesting input on any administrative
requirements that were duplicative or outdated. Over 146 comment
submissions, with 100 unique recommendations, were received. The NRC
evaluated comments per criteria approved in SRM-COMSECY-18-0027,
``Staff Requirements--COMSECY-18-0027--Evaluation Criteria for
Retrospective Review of Administrative Regulations (RROAR)''
(ML19281C697). On June 1, 2021, the NRC staff summarized the evaluation
in SECY-21-0056, ``Evaluation of Stakeholder Input on Retrospective
Review of Administrative Requirements'' (ML21013A026), screening 54
recommendations into a rulemaking. The NRC staff coordinated public
meetings on June 30 and August 19, 2021, to discuss the results of the
RROAR evaluation. On September 8, 2021, NEI provided a letter to the
NRC stating the limited value in continuing the RROAR effort
(ML21256A149). On December 20, 2021, the NRC staff submitted rulemaking
plans in SECY-21-0110, ``Rulemaking Plans to Implement the Results of
the Retrospective Review of Administrative Requirements Evaluation''
(ML21124A116) requesting Commission approval to initiate rulemaking.
The NRC staff proposed two rulemakings, a streamlined approach (e.g.,
no regulatory basis) for administrative changes and a standard approach
for more significant changes. Additional information is available on
the NRC public website at <a href="https://www.nrc.gov/about-nrc/regulatory/rulemaking/retrospective-review-admin-reqmnts.html">https://www.nrc.gov/about-nrc/regulatory/rulemaking/retrospective-review-admin-reqmnts.html</a>. This proposed
rulemaking would include revisions that incorporate or build on these
RROAR activities, including Revision of Administrative Requirements
(see SECY-21-0110, ML21124A116) and the development of the proposed
changes to Sec. 50.72 in this proposed rule. In large part, the
proposed revisions would align with E.O.'s 14300, 14192, and 14270,
``Zero-Based Regulatory Budgeting To Unleash American Energy'' by
reducing regulatory burden through the modification or elimination of
outdated or redundant administrative requirements identified in RROAR
and other reporting requirements. Consistent with E.O. 14270, this
rulemaking would remove unnecessary reporting obligations, streamlining
information submission, recordkeeping, and reporting processes without
compromising the NRC's mission.

IV. Discussion

This discussion section has been divided into subsections to better
present information on this proposed rule. The following sections
correspond to the associated sections of the same title in Section III,
``Background,'' of this document. Each section discusses what action
the NRC is proposing and whom the action would affect and how they
would be affected.

10 CFR Parts 21, 50, and 53--10 CFR Part 21 Clarifications

A. 10 CFR Part 21, and Sec. Sec. 50.55(e), and 53.605 Alignment
Section 50.55(e) has been modified over the last 30 years to become
more aligned with 10 CFR part 21. Incorporating the two remaining
differences from Sec. 50.55(e) into 10 CFR part 21 would allow for the
deletion of Sec. 50.55(e). There are no identified disadvantages
associated with combining these two regulations. The following proposed
changes would remove redundant regulation and clarify regulatory
requirements.
This proposed rule would delete Sec. 21.2(b) and renumber the
subsequent sections appropriately. This proposed rule would add Sec.
21.2(a)(5) to address 10 CFR part 21 applicability to a construction
permit, early site permit, and or manufacturing license and incorporate
language from Sec. 21.2(b).
In order to incorporate Sec. 50.55(e) requirements into 10 CFR
part 21, this proposed rule would add a new Sec. 21.21(a)(3)(iii) with
the following statement from Sec. 50.55(e): Undergoes any significant
breakdown in any portion of the quality assurance program conducted
under the requirements of appendix B to 10 CFR part 50 which could have
produced a defect in a basic component. These breakdowns in the quality
assurance program are reportable whether or not the breakdown actually
resulted in a defect in a design approved and released for
construction, installation, or manufacture.
These proposed rule changes would add a new Sec. 21.21(d)(1)(iii)
and (d)(2) to include requirements for the holder of a facility
construction permit subject to 10 CFR part 21, combined license, or
manufacturing license, who obtains

[[Page 60721]]

information reasonably indicating that the quality assurance program
has undergone any significant breakdown, to notify the Commission of
the breakdown in the quality assurance program through a director or
responsible officer or designated person.
This proposed rule would add a new Sec. 21.51(b)(1)-(3). These
paragraphs would state that the holder of a construction permit,
combined license, and or manufacturing license must prepare and
maintain records necessary to accomplish the purposes of Sec. 21.51.
Specifically, this requirement would include retaining records of
evaluations of all deviations and failures to comply under Sec.
21.21(a)(1) for the longest of 10 years from the date of the
evaluation; 5 years from the date that an early site permit is
referenced in an application for a combined license; or 5 years from
the date of delivery of a manufactured reactor.
This proposed rule would add a new Sec. 21.51(b)(4) for suppliers
of basic components for holders of construction permits, combined
licenses, and manufacturing licenses. This section would state that
such suppliers of basic components must retain a record of all
notifications sent to affected licensees or purchasers under Sec.
21.21(c) for a minimum of 10 years following the date of the
notification, and a record of the facilities or other purchasers to
whom basic components or associated services were supplied for a
minimum of 15 years from the delivery of the basic component or
associated services.
This proposed rule would incorporate in 10 CFR part 21 applicable
portions of the definition of ``basic component'' from 10 CFR part 50.
Specifically, the language in paragraph (2) of the definition of
``basic component'' in Sec. 50.2 would be added to the definition of
``basic component'' in Sec. 21.3, ``Definitions,'' as a sixth item to
include applicability to a construction permit holder, consistent with
Sec. 50.55(e).
This proposed rule would delete Sec. Sec. 50.55(e) and 53.605 in
their entirety. The proposed change of relocating specific Sec. Sec.
50.55(e) and 53.605 requirements to 10 CFR part 21 would not impact
licensees as compliance with the current Sec. Sec. 50.55(e) and 53.605
would be sufficient to comply with the requirements proposed to be
relocated. Additionally, there are currently no existing 10 CFR part 53
licensees.
B. Commercial Grade Dedication Definition for 10 CFR Parts 21, 52, and
53
This proposed rule would modify the definitions of ``commercial
grade item'' and ``dedication'' in Sec. 21.3 to include references to
10 CFR parts 52 and 53. This change would address (1) concerns raised
after the 2007 10 CFR part 52 final rule, in which the NRC recognized
that conforming changes to definitions in 10 CFR part 21 were not
included to address 10 CFR part 52; and (2) the recent issuance of the
10 CFR part 53 final rule.
This proposed rule change for including references to 10 CFR parts
52 and 53 in the 10 CFR part 21 definitions of ``commercial grade
item'' and ``dedication'' would not impact licensees and other
stakeholders that are already implementing the commercial grade
dedication acceptance process.
C. Executive Order 14267, Reducing Anti-Competitive Regulatory Barriers
This proposed rule would revise the Sec. 21.3 definition of
``dedicating entity'' to address E.O. 14267. The proposed rule would
also delete the word ``third-party'' with respect to who can perform
the dedication, which has been determined to be anticompetitive,
because it could limit competition among qualified suppliers.
The proposed rule would also clarify that a 10 CFR part 50,
appendix B or appendix T compliant quality assurance program and 10 CFR
part 21 program are needed when performing the commercial grade
dedication process. In addition, the proposed rule would provide
conforming changes to address the issuance of 10 CFR part 53 and the
proposed rule to add 10 CFR part 50, appendix T. Revising the
definition of ``dedicating entity'' would not impact licensees and
other stakeholders that are already implementing the commercial grade
dedication acceptance process. In addition, a clarification was added
to the definition of ``dedicating entity'' to ensure that anti-
competitive regulatory barriers are not introduced for fuel facilities.
Fuel facilities, uranium enrichment and fuel fabrication facilities,
plutonium processing and fuel fabrication plants, and spent fuel
reprocessing facilities, that are licensed under 10 CFR part 70 and are
required to develop an integrated safety analysis in accordance with
subpart H to 10 CFR part 70, may have a dedicating entity perform
dedication so long as the organization has a quality assurance program
compliant with the licensee's approved configuration management program
or management measure program, respectively. The proposed rule language
is consistent with current licensee practices and aligns with the
existing text in 10 CFR part 21 related to dedication for nonreactors.
D. Clarify ``Basic Component'' Definition
This proposed rule would modify the definition of ``basic
component'' in Sec. 21.3 to provide clarity to uranium enrichment,
fuel fabrication, fuel conversion, plutonium processing and fuel
fabrication plants or spent fuel processing facilities that are
required, by regulation or license condition, to develop an integrated
safety analysis in accordance with subpart H to 10 CFR part 70. For
fuel facilities, this modification would focus the scope of commercial
grade dedication to sole engineered items relied on for safety. Spent
fuel reprocessing facilities and plutonium processing and fuel
fabrication plants would also focus on sole engineered items relied on
for safety and, additionally, include all engineered equipment that is
relied on to prevent or mitigate high-consequences to individuals
located beyond the controlled area boundary from a release of fission
products and transuranics. While 10 CFR part 70 facilities are clearly
within the scope of 10 CFR part 21, the NRC had previously received
feedback that the definition of ``basic component'' in Sec. 21.3, was
difficult to interpret as it was written to have applicability to
multiple other types of facilities and activities and did not offer
sufficient specificity for such varied activities. In contrast, the
power reactor definition of a basic component is specific to power
reactor terminology. As a result, the fuel facility industry
communicated through the years that they had difficulty in applying the
definition as written, as made evident by numerous facility-specific
definitions or exemptions. The revised definition of basic component
would clarify the connection between subpart H to 10 CFR part 70, and
10 CFR part 21 using insights gained from years of operating
experience, the current regulatory framework,\4\ and current industry
procurement practices.
---------------------------------------------------------------------------

\4\ Subpart H to 10 CFR part 70 was issued on September 18, 2000
(65 FR 56211).
---------------------------------------------------------------------------

The NRC is proposing to take a risk-informed approach in the
definition of basic component as applied to fuel facilities. The
licensing basis for fuel facilities is well suited to taking such a
risk-informed approach, as licensees that are subject to 10 CFR part
70, subpart H, are required to perform an integrated safety analysis.
The integrated safety analysis designates items relied on for safety,
consisting of passive engineered controls, active engineered controls,
administrative controls, and enhanced administrative

[[Page 60722]]

controls (or a combination of these), to prevent or mitigate the
consequences of events that could cause the performance requirements to
be exceeded without the application of the items relied on for safety.
These items relied on for safety are subject to requirements for
configuration management, maintenance, training, reporting, and other
quality assurance elements to ensure they are reliable and available to
function as designed. In many cases, licensees have multiple items
relied on for safety to independently prevent or mitigate a specific
safety hazard. However, in some cases, licensees may have only one
`sole' engineered item relied on for safety. In these cases, the NRC
proposes that the sole item relied on for safety be designated a basic
component.
By providing a new definition of basic components that would be
specific to fuel facilities, plutonium processing and fuel fabrication
plants or spent fuel reprocessing facilities, the NRC would provide
clarity about the components that would be managed under a commercial
grade dedication program and would need to be evaluated for deviations
or defects. This definition, for fuel facilities, would focus
dedication activities to components that are sole engineered items
relied on for safety for preventing or mitigating intermediate-
consequence events and high-consequence events including accidental
criticalities. For spent fuel reprocessing facilities and plutonium
processing and fuel fabrication plants, an additional criterion to this
definition is proposed that would focus dedication activities on items
relied on for safety that are necessary to prevent or mitigate hazards
to members of the public located beyond the controlled area boundary
unique to such facilities.
If a loss or degradation of any item relied on for safety results
in failure to meet the performance requirements of Sec. 70.61, the
licensee is required to report the failure to the NRC. The new
definition does not change the reporting requirements of either 10 CFR
part 21 or 10 CFR part 70.
The definition of ``basic component'' would be modified to provide
consistency with the definition of safety-related SSCs under 10 CFR
part 53. Specifically, the changes to the definition of ``basic
component'' made by the 10 CFR part 53 rule are inconsistent with the
scope of what structures, systems, and components are considered
safety-related in 10 CFR part 53. Therefore, this revision would
correctly align the definition of ``basic component'' with the
technology-inclusive framework of 10 CFR part 53.
Additionally, the definition of ``basic component'' would be
modified to allow items designed and manufactured under a quality
assurance program that is complying with the proposed appendix T to 10
CFR part 50 to be included in this definition. This change would not
impact licensees as compliance with the proposed appendix T would be a
voluntary alternative to appendix B to 10 CFR part 50.
E. Clarify That Commercial Grade Dedication Is Not Applicable to
Nuclear Material Users
This proposed rule would remove the references to 10 CFR part 30
from the definitions of ``commercial grade item'' and ``dedication'' in
Sec. 21.3. This change would clarify that broad scope and specific
licensees (e.g., radiography, medical licensees, irradiators, and well
logging) with licenses issued under 10 CFR parts 30, 31, 34, 35, 36,
and 39, would not need to dedicate their basic components using the
commercial grade dedication process. This proposed change would make
the regulatory language consistent with industry practice. The NRC has
not provided guidance to nuclear material users on the commercial grade
dedication process, and enforcement of this portion of 10 CFR part 21
is of low safety significance. This proposed change would correct the
regulation to ensure that nuclear material users continue to report
defects on basic components but are not required to commercially
dedicate them.
F. Remove 10 CFR Part 76 References
This proposed rule change would also remove references to 10 CFR
part 76, which applied to gaseous diffusion plants. The proposed
deletion of references to 10 CFR part 76 would not impact licensees and
other stakeholders because the facilities subject to 10 CFR part 76 are
no longer in operation and were transitioned to the regulatory
oversight of the U.S. Department of Energy for decommissioning. The
administrative removal of references to 10 CFR part 76 from 10 CFR part
21 would not impact NRC resources or external stakeholders. The NRC
issued a direct final rule on December 3, 2025 (90 FR 55621), that
amended various regulations to insert a conditional sunset date in
response to E.O. 14270. This direct final rule added the conditional
sunset date to 10 CFR part 76.
G. Update Notification Methods
This proposed rule change would update Sec. 21.21(d)(3)(i) by
replacing facsimile information with the NRC Headquarters Operations
Officer (HOO) email address: <a href="/cdn-cgi/l/email-protection#7a121515541215193a140819541d150c">[email&#160;protected]</a>. The proposed change would
replace initial notification by facsimile with notification by email to
the NRC HOO email address: <a href="/cdn-cgi/l/email-protection#177f7878397f78745779657439707861">[email&#160;protected]</a> as the preferred method,
which would not impact licensees and other stakeholders as some are
already implementing notification using the NRC HOO email.

10 CFR Part 50--Reporting Requirements for Nonemergency Events for
Nuclear Power Plants

A. Purpose and Scope
This proposed rule would revise Sec. 50.72(b) to eliminate
reporting requirements associated with certain nonemergency events and
allow licensees to use alternative, technology-inclusive methods for
notifying the NRC of certain nonemergency events at commercial nuclear
power plants. This proposed rule would better align the reporting
requirements with the current needs of the agency and licensees by
eliminating those reports that are not necessary for the NRC to have
reasonable assurance of adequate protection of the public health and
safety, that do not require prompt NRC action, or that can be tracked
using other existing agency processes. These proposed changes would
help ensure that the agency devotes its resources to those events for
which the agency needs to take prompt action while maintaining
appropriate NRC and stakeholder awareness of these events. This
proposed rule also would allow for the submission of certain
nonemergency event reports by methods other than the Emergency
Notification System (ENS). Accordingly, this proposed rule would result
in a reporting process that maintains openness, reduces regulatory
burden by focusing requirements on issues important to safety
warranting prompt NRC actions, and addresses advances in communication
technology. The safety significance of the reported issues is the
primary decision criteria for the proposed changes (i.e., likelihood of
the event, consequences of the event, and defense-in-depth
considerations). Public awareness of nonemergency events by itself is
not a significant factor for this proposed rule.
As part of its evaluation of each of the reporting requirements in
Sec. 50.72(b), the NRC considered the Principles of Good Regulation,
the Be riskSMART framework, and the internal NRC evaluation and
communications process for nonemergency event notifications (discussed
in section 2.4 of the regulatory basis). The Principles of Good
Regulation include independence,

[[Page 60723]]

openness, efficiency, clarity, and reliability and ensure that the
agency focuses on its safety and security mission while appropriately
balancing the interests of the NRC's stakeholders, including the public
and licensees. The Be riskSMART framework provides a process to ensure
the consistent application of risk information across different NRC
program areas. Use of the Principles of Good Regulation and the Be
riskSMART framework enabled the NRC to develop standard criteria by
which to assess each of the reporting requirements. Under these
criteria, the relative frequency of reporting of a specific event
notification is not the sole factor in the justification for
elimination of a requirement. The inclusion of the numbers of reactive
inspections initiated due to event notifications provides some valuable
risk insights (i.e., unless deterministic criteria are identified, the
decision to initiate a reactive inspection would be primarily based on
conditional core damage probability). For some categories of reporting,
the relative frequency of reports is low, challenging the basis
described in PRM-50-116 about the administrative burden on licensees.
Conversely, for some categories of reporting where there were a
relatively greater number of reports, there were few of these events
that tripped the risk threshold to initiate a reactive inspection, and
most follow-up was conducted by resident inspectors under the baseline
inspection program. Using the Be riskSMART framework, the data on
frequency of event notification reports and the historical NRC response
(i.e., reactive inspection versus baseline inspection follow-up) was
used to help inform the risk of eliminating these reports in the
context of the original bases for the 4- and 8-hour reports. The NRC's
application of these criteria is detailed in section 3.0 of the
regulatory basis.
In this proposed rule, the NRC is not proposing changes to the
following nonemergency reporting requirements:
a. Section 50.72(b)(1)--notification to the NRC within 1 hour of
any deviation from the plant's technical specifications authorized by
10 CFR 50.54(x), if not already reported as part of a declared
emergency;
b. Section 50.72(b)(2)(i)--notification to the NRC within 4 hours
of the initiation of any nuclear plant shutdown required by the plant's
technical specifications;
c. Section 50.72(b)(2)(iv)(B)--notification to the NRC within 4
hours of an event or condition that results in actuation of the reactor
protection system when the reactor is critical, except when the
actuation results from and is part of a pre-planned sequence during
testing or reactor operation;
d. Section 50.72(b)(3)(ii)(A)--licensees to submit a report to the
NRC within 8 hours of any event or condition that results in the
condition of the nuclear power plant, including its principal safety
barriers, being seriously degraded;
e. Section 50.72(b)(3)(ii)(B)--licensees to submit a report to the
NRC within 8 hours of any event or condition that results in the
condition of the nuclear power plant being in an unanalyzed condition
that significantly degrades plant safety; and
f. Section 50.72(b)(3)(xiii)--notification to the NRC of an event
that results in a major loss of offsite communications capability,
offsite response capability, or emergency assessment capability.
Eliminating or delaying these reports would unacceptably degrade
the NRC's ability to perform oversight and incident response functions
for the potential significance or safety issue associated with those
events. The NRC is also not proposing changes to the emergency
reporting requirements found in appendix E, ``Emergency Planning and
Preparedness for Production and Utilization Facilities,'' to 10 CFR
part 50; the reporting requirements of Sec. 50.73; or other event
reporting requirements, such as those contained 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 10 CFR part 70, ``Domestic Licensing of Special Nuclear Material.''
B. Proposed Changes to Sec. 50.72
The NRC proposes to amend the following nonemergency event
reporting requirements in Sec. 50.72. Eliminating these event reports
would not be an unacceptable degradation of situational awareness
because a delay in NRC awareness for the scope of these events and
otherwise maintaining awareness via routine oversight processes would
be commensurate with the safety significance of issues associated with
these event reports. This proposed rule would eliminate the following
reporting requirements.
a. Section 50.72(b)(2)(iv)(A)--the requirement to notify the NRC of
an event that results or should have resulted in an ECCS discharge into
the reactor coolant system as a result of a valid signal except when
the actuation results from and is part of a pre-planned sequence during
testing or reactor operation;
b. Section 50.72(b)(2)(xi)--the requirement to notify the NRC of an
event or situation related to the health and safety of the public or
onsite personnel, or protection of the environment, for which a news
release is planned or notification to other government agencies has
been or will be made;
c. Section 50.72(b)(3)(iv)(A)--the requirement to notify the NRC of
an event or condition that results in valid actuation of specified
safety systems;
d. Section 50.72(b)(3)(v)--the requirement to notify the NRC of an
event or condition that could have prevented fulfillment of the safety
function of structures or systems needed to shut down and maintain the
reactor in a safe condition, remove residual heat, control the release
of radioactive material, or mitigate the consequences of an accident;
and
e. Section 50.72(b)(3)(xii)--the requirement to notify the NRC of
an event requiring the transport of a radioactively contaminated person
to an offsite medical facility for treatment.
This proposed rule would also add new paragraph (a)(6) to Sec.
50.72 to enable licensees to submit certain nonemergency event reports
via methods other than the ENS.
i. Proposed Changes to Sec. 50.72(b)(2)(iv)(A)
In Sec. 50.72(b)(2)(iv)(A), the NRC currently requires that each
nuclear power reactor licensee notify the NRC of an event that results
or should have resulted in a discharge of the ECCS into the reactor
coolant system resulting from a valid signal unless the actuation
resulted from and was part of a pre-planned sequence during testing or
reactor operation. The Sec. 50.72(b)(2)(iv)(A) report is required to
be made within 4 hours of occurrence of such an event.
The NRC proposes to eliminate this reporting requirement because
these reports are infrequent, occurring 37 times across the entire
nuclear power plant fleet between 2011 and 2021, and have not required
prompt agency actions as discussed in section 14.2 of the regulatory
basis. Agency records indicate that the NRC has never activated its
response plan in response to these events absent an emergency
declaration. Most instances where the underlying events result in a
notification under this requirement will happen when the RPS is in
service. ECCS actuations that also result in RPS actuation signals
would still be captured by the RPS actuation notification requirement
in the current Sec. 50.72(b)(2)(iv)(B). Further, if the

[[Page 60724]]

underlying condition that resulted in the ECCS actuation meets the
threshold for an emergency declaration as required by a licensee's
emergency plan, such conditions would still be reportable within 1 hour
under Sec. 50.72(a)(3).
Finally, there may be some low-risk conditions where ECCS
actuations occur that do not result in RPS actuations or meet
licensees' pre-defined emergency declaration conditions. These low-risk
actuations would still be captured in the LER reporting requirement in
Sec. 50.73(a)(2)(iv), which would continue to ensure openness by
making information about these events publicly available. As those
conditions are of low risk, they would not result in the need for
prompt agency action, and reporting only via LER would be appropriate.
Inspectors would likely identify and may follow up on these events
during their normal oversight duties well before an LER was issued,
depending on the safety significance of the event. Further discussion
of the factors and options considered by the NRC can be found in
section 14.0 of the regulatory basis.
ii. Proposed Changes to Sec. 50.72(b)(2)(xi)
In Sec. 50.72(b)(2)(xi), the NRC requires that licensees notify
the NRC of any event or situation, related to the health and safety of
the public or onsite personnel, or protection of the environment, for
which a news release is planned or notification to other government
agencies has been or will be made. Such an event may include an onsite
fatality or inadvertent release of radioactively contaminated
materials. The report is required to be made within 4 hours of
occurrence of such event.
The NRC proposes to eliminate this reporting requirement because
these events are generally not related to nuclear plant safety and,
therefore, do not pose a threat to public health and safety. Agency
records indicate that the NRC has never initiated an inspection or
investigation in response to these events. Removing this requirement
would not affect the NRC's ability to execute oversight, event
response, or other regulatory functions.
The NRC and the public may first become aware of issues, such as
fatalities or contamination, from news and social media. In addition,
public or NRC situational awareness of significant items would continue
to occur through the normal regional baseline inspections and
subsequent inspection reports that already address the underlying
condition resulting in offsite nonemergency event notifications. As an
example, resident inspectors routinely sample and review condition
reports generated on a daily basis under the corrective action program.
These condition reports capture situations that would require reporting
by the current criteria proposed to be eliminated by this rulemaking.
The NRC would remain properly informed of these events and be able to
communicate them to the public via quarterly baseline inspection
reports, if appropriate. Responses to any public inquiries could be
developed and coordinated through the NRC's Office of Public Affairs.
iii. Proposed Changes to Sec. 50.72(b)(3)(iv)
In Sec. 50.72(b)(3)(iv)(A), the NRC requires that nuclear power
reactor licensees notify the NRC of an event or condition that results
in valid actuation of any of the systems listed in Sec.
50.72(b)(3)(iv)(B) except when the actuation results from and is part
of a pre-planned sequence during testing or reactor operation. The
report must be made within 8 hours of the occurrence of such an event.
Section 50.72(b)(3)(iv)(B) lists the systems subjected to the reporting
requirements in Sec. 50.72(b)(iv)(A), and the NRC is proposing to
delete this list as a conforming change. The systems listed in Sec.
50.72(b)(3)(iv)(B) are the following: reactor protection system
including reactor scram and reactor trip; general containment isolation
signals affecting containment isolation valves in more than one system
or multiple main steam isolation valves; ECCS for pressurized water
reactors including high-head, intermediate-head, and low-head injection
systems and the low pressure injection function of residual (decay)
heat removal systems; ECCS for boiling water reactors including high-
pressure and low-pressure core spray systems, high-pressure coolant
injection system, and low pressure injection function of the residual
heat removal system; boiling water reactor core isolation cooling
system, isolation condenser system, and feedwater coolant injection
system; pressurized water reactor auxiliary or emergency feedwater
system; containment heat removal and depressurization systems,
including containment spray and fan cooler systems; and emergency ac
electrical power systems, including emergency diesel generators (EDGs),
hydroelectric facilities used in lieu of EDGs at the Oconee Station,
and boiling water reactor dedicated Division 3 EDGs.
The NRC proposes to eliminate this reporting requirement because
these reports have not required prompt agency oversight or enforcement
action within a day as discussed in Section 19.2 of the regulatory
basis. The routine follow-up of plant issues already performed by
resident inspectors is sufficient to ensure that the NRC takes
appropriate near-term actions in response to these events.
Historically, the NRC has not needed to take prompt actions for event
notifications reported under this criterion outside of routine event
response, thus, retaining this criterion is not warranted. Formal
notification through LERs under Sec. 50.73(a)(2)(iv) would provide the
NRC with more detailed information on the event and inform longer-term
actions to respond to such events. Further discussion of the factors
and options considered by the NRC can be found in section 19.0 of the
regulatory basis described in the ``Part 50--Reporting Requirements for
Nonemergency Events for Nuclear Power Plants'' section of this
document.
iv. Proposed Changes to Sec. 50.72(b)(3)(v)
In Sec. 50.72(b)(3)(v), the NRC requires that nuclear power
reactor licensees notify the NRC of an event or condition that, at the
time of discovery, could have prevented the fulfillment of the safety
function of structures or systems that are needed to shut down the
reactor and maintain it in a safe shutdown condition, remove residual
heat, control the release of radioactive material, or mitigate the
consequences of an accident. The report must be made within 8 hours of
the occurrence of such an event.
The NRC proposes to eliminate this reporting requirement because
these reports have been infrequent and rarely require prompt agency
oversight or enforcement action. The routine follow-up of plant issues
already performed by resident inspectors is sufficient to ensure that
the NRC takes appropriate near-term actions in response to these
events. Formal notification through LERs under Sec. 50.73(a)(2)(iv)
would provide the NRC with more detailed information on the event and
inform longer-term actions to respond to such events. Further
discussion of the factors and options considered by the NRC can be
found in section 20.0 of the regulatory basis.
v. Proposed Changes to Sec. 50.72(b)(2)(xii)
In Sec. 50.72(b)(3)(xii), the NRC requires that nuclear power
reactor licensees notify the NRC of any event requiring the transport
of a radioactively contaminated person to an offsite medical facility
for treatment. This report is required to be made within 8 hours of the
occurrence of such an event.
The NRC proposes to eliminate this reporting requirement because
events

[[Page 60725]]

covered by Sec. 50.72(b)(3)(xii) are generally not related to plant
safety and licensees and offsite emergency responders have arrangements
and agreements in place to safely transport contaminated individuals
without compromising public health and safety. Specifically, the
regulations in paragraph (b)(12) of Sec. 50.47, ``Emergency plans,''
and appendix E to 10 CFR part 50 require licensees to have arrangements
and agreements in place for transportation and medical services for
radiologically contaminated individuals, which includes notification of
the offsite medical facilities. The regulations in paragraph (a)(12) of
44 CFR 350.5, ``Criteria for review and approval of State and local
radiological emergency plans and preparedness,'' and the Federal
Emergency Management Agency's (FEMA) Radiological Emergency
Preparedness Program Manual provide policies and guidance to offsite
response organizations, including to have these arrangements with
licensees as well. The NRC regulations in Sec. 50.47(b)(14) and
appendix E to 10 CFR part 50 and the FEMA regulations in 44 CFR
350.5(a)(14) also contain training provisions for licensees and offsite
emergency responders, respectively. Therefore, eliminating the
reporting requirement in Sec. 50.72(b)(3)(xii) would not pose a threat
to public health and safety.
Agency records indicate that the NRC has never initiated an
inspection or investigation in response to these events and would still
be able to successfully execute its oversight, event response, and
regulatory functions. Further discussion of the factors and options
considered by the NRC can be found in section 21.0 of the regulatory
basis.
vi. Proposed Changes to Sec. 50.72(a)
The NRC is proposing to allow licensees to use alternative methods
to notify the NRC of nonemergency events listed in Sec. 50.72(b)(2)-
(3). Currently, each operating nuclear power reactor licensee is
required, in part, under Sec. 50.72(a)(1) to notify the NRC
Headquarters Operations Center (HOC) through the ENS of the declaration
of any of the emergency classes specified in the licensee's approved
emergency plan and nonemergency events specified in Sec. 50.72(b). The
ENS lines are direct access telephone lines that bypass switching in
the telephone service's local central office. These lines are installed
in the licensee control rooms as well as their technical support center
and emergency operations facility. If the ENS is inoperative, Sec.
50.72(a)(2) requires the licensee to make the required emergency
notifications through a commercial telephone service, another dedicated
telephone system, or any other method enabling a report as soon as
practical to the HOC at the telephone numbers specified in appendix A
to 10 CFR part 73, ``Physical Protection of Plants and Materials.''
Although use of NRC Form 361, ``Reactor Plant Event Notification
Worksheet,'' is voluntary, it provides a standard structure for
licensees to communicate to the HOC the information required by Sec.
50.72.
In the early 1980s, when the NRC first required and implemented the
ENS, telephones were the only widely available means of communication
available between the NRC and its licensees. Since then, several other
means of communication have become widely available, including
facsimile, email, text messaging, instant messaging, and online
portals. Licensees have used facsimile or other electronic means to
submit a completed version of Form 361 to the NRC with specific event
information as a supplement to their telephone reporting required by
Sec. 50.72(a)(1). Often, the completed version of NRC Form 361 would
be sufficient to notify the NRC of the event. However, licensees must
continue to use the ENS to call the HOC because of the requirement to
do so in Sec. 50.72(a)(1). In cases for which NRC Form 361 would be
sufficient, the mandatory ENS phone call would be unnecessary for many
nonemergency reports.
The NRC is proposing to add a new Sec. 50.72(a)(6) to allow
licensees to submit nonemergency event reports required under Sec.
50.72(b)(2)-(3) using other methods as an alternative to using the ENS,
provided the licensee submits the report to the NRC HOC within the
required timeframe and confirms the NRC's receipt of the report. For
example, the NRC offers certain persons an online method of submitting
reports to the NRC, although licensees are still required by regulation
to communicate all events directly to the HOC through the ENS.
Beginning in 2022, the NRC began allowing properly vetted licensee
personnel to provide NRC Form 361 through an online portal. Upon
submittal, the information is sent to the HOC for the Headquarters
Operations Officers for follow up with the licensee, if needed, and
process to the NRC's operating experience team (nuclear power plants
only) for data retention, tracking, and trending. Under this proposed
rule, licensees using this online portal for Sec. 50.72(b)(2) or (3)
nonemergency event reports would need only to confirm NRC receipt of
the report and would not need to use the ENS.
The NRC would not allow licensees to submit nonemergency event
reports required under Sec. 50.72(b)(1) using methods permitted by
proposed Sec. 50.72(a)(6). The event listed in Sec. 50.72(b)(1)--
deviation from the plant's technical specifications, if not already
reported as part of a declared emergency--is required to be reported
within 1 hour of its occurrence because the event is ongoing, and
actions are being taken to address immediate safety or security
concerns. The nature of these reports is such that prompt actions or
follow-up by the NRC may be warranted commensurate with safety
significance. The HOC is the primary center of communication and
coordination among the NRC, its licensees, and other stakeholders for
operating events and needs to be made aware of these events within 1
hour. Event notifications required by Sec. 50.72(b)(1) must be made
through the ENS so the HOC promptly receives them.
C. NRC Situational Awareness and Public Openness
For events associated with reporting criteria that are proposed for
elimination, NRC situational awareness would continue under the normal
inspection activities as part of the Reactor Oversight Process (ROP).
These events would be treated as other plant conditions of similar
safety significance are treated and would be subject to each licensee's
informal communication processes and the NRC's baseline sampling
process that resident inspectors apply to all reported plant
conditions. Resident inspection staff would continue to perform
baseline inspections of selected conditions and document findings in
inspection reports regardless of reportability under the nonemergency
event notification criteria. As an example, resident inspectors
routinely review daily plant condition reports under the corrective
action program and through their routine plant status activities. Those
baseline inspection activities capture conditions of potential public
health and safety concern occurring at licensee sites, including
situations that would be reported to offsite organizations. The
proposed changes are not intended to increase the workload of resident
inspectors or require resident inspectors to gather, verify, or
communicate information beyond what is done through existing inspection
activities.
Many ROP-related activities include actions to inform the NRC of
items of interest within a sufficient timeframe, such as those
specified in inspection

[[Page 60726]]

procedures, routine regional status meetings, and frequent informal
communications between inspectors and regional management. For
significant items discovered during these activities, regional staff
perform reactive inspection evaluations in accordance with MD 8.3,
``NRC Incident Investigation Program.'' The MD 8.3 process is the NRC's
primary internal tool for ensuring that significant operational events
are investigated in a timely, objective, systematic, and technically
sound manner. These activities, which occur independently of
nonemergency event notifications, ensure that the NRC would remain
informed, act in a timely manner, and communicate items of interest
internally.
Although the proposed approach would eliminate prompt notifications
for some events that have no or very low safety or regulatory
significance, the NRC would maintain sufficient NRC situational
awareness of these events, as appropriate, via regional baseline and
emergency preparedness inspection activities, including available
performance indicators. Notification of individual events to NRC
Headquarters technical points of contact and the public may be delayed
or not occur for a subset of issues. For example, a situation that
results in an ECCS system actuation but not an associated RPS actuation
is currently reportable under either a 4-hour or 8-hour report
depending on the circumstances. Such a report is shared internally with
NRC Headquarters technical points of contact and management within 24
hours of receipt for situational awareness. As proposed in this
rulemaking, this event might not be formally reported to the NRC until
a required LER is made 60 days after the event or not at all. However,
regional inspection staff and management would become aware of these
ECCS events within several days through performing routine baseline
inspection activities. Other NRC technical points of contact and
management may not be promptly notified for purposes of situational
awareness but may become aware through other routine communication
processes. As another example, critical emergency preparedness
assessment capabilities are addressed in licensee EAL schemes, and a
major loss of capabilities would result in a 1-hour emergency event
notification per Sec. 50.72(a)(3), thereby providing the NRC with
sufficient situational awareness.
Openness and public awareness would be maintained under the normal
regional baseline inspections and subsequent inspection reports that
already address situations resulting in offsite notifications. The
release of public information on some plant events may be delayed, or
not occur, for some events that have low safety or regulatory
significance. As an example, resident inspectors routinely review daily
plant condition reports generated under the corrective action program.
This review includes the underlying conditions that are currently
required to be reported to the NRC but would not be immediately
reported under this proposed rule, thereby ensuring the NRC would
remain informed and able to communicate issues of safety or regulatory
significance to the public via quarterly baseline inspection reports.
If a nonemergency report is not generated and interest from
external stakeholders exists, then the NRC would consider communication
of event information to stakeholders, where appropriate, following the
current processes for communicating items of interest. For issues of
public interest, responses could be developed and coordinated through
the NRC's Office of Public Affairs. Internal interest could be managed
using existing processes and regular operating experience briefs.
In summary, the inspection program would ensure sufficient openness
and sharing of information with the public for significant issues
subject to LER reporting criteria and, for items where findings are
identified, through inspection reports. However, this information
sharing would take longer than for the current event notification
criteria and may not occur for items that have no associated LER
reporting criteria (e.g., major loss of communications) and for which
regulatory findings are not identified.

10 CFR Parts 20, 50, 52, and 53--Decommissioning Licensing Enhancements

New and advanced reactor technologies in a variety of sizes,
geographic locations, technical complexities, and licensed footprints
may be licensed under the long-standing frameworks in 10 CFR parts 50
and 52 or the recently published 10 CFR part 53. As the NRC prepares to
license reactors that differ from the traditional, large light-water
reactor designs that the United States has seen for decades, the NRC
intends to establish and maintain regulatory frameworks that are risk-
informed and technology-inclusive. Given the number of options
regarding regulatory frameworks that an applicant may pursue to license
a specific reactor technology, the NRC is proposing to apply regulatory
enhancements to decommissioning licensing requirements across all three
frameworks (i.e., 10 CFR parts 50, 52, and 53). These enhancements are
described in the following subsections.
A. Expansion of Partial Site Release Requests
The NRC proposes to add new Sec. 52.111, ``Release of part of a
power reactor facility or site for unrestricted use,'' as a specific
provision for release of a part of a site for unrestricted use before
license termination for part 52 licensees. The proposed Sec. 52.111
would be nearly identical to the text in Sec. 50.83, with specific
applicability to holders of an operating license or combined license
under 10 CFR part 52. The rule language from the existing Sec. 50.83
and the regulations referenced within it are otherwise already
applicable for 10 CFR part 52 facilities and no other changes to the
language from the existing Sec. 50.83 would be needed. The proposed
changes are anticipated to provide 10 CFR part 52 licensees with a
clearer and more efficient path towards releasing portions of their
site prior to license termination and will align the regulations in 10
CFR part 52 with those in 10 CFR parts 50 and 53.
Minor conforming changes are also proposed for Sec. 20.1401,
``General provisions and scope,'' in paragraphs (a) and (c). The
proposed change to Sec. 20.1401(a) would add the release of a part of
a facility or site for unrestricted use in accordance with proposed
Sec. 52.111 to the list of facilities that the criteria in part 20,
subpart E, ``Radiological Criteria for License Termination,'' apply to.
The proposed change to Sec. 20.1401(c) would limit the circumstances
in which the Commission could require additional cleanup after release
of part of a facility or site for unrestricted use in accordance with
proposed Sec. 52.111 to those instances where based on new information
it is determined that the criteria of part 20 subpart E were not met
and residual radioactivity remaining at the site could result in
significant threat to public health and safety.
B. Reducing Regulatory Requirements for Prompt Decommissioning of Power
Reactors
The NRC is proposing to expand the options by which nuclear power
reactors may terminate their license. Because of the range of
commercial and industrial utilization and production facilities that
may become licensed under 10 CFR part 50, 10 CFR part 52, or 10 CFR
part 53, the NRC is proposing to enhance the decommissioning licensing
process to accommodate the unique aspects of a variety of facilities.

[[Page 60727]]

Currently, a nuclear power reactor licensee must submit a PSDAR within
2 years of permanent shutdown. The licensee may perform dismantlement
and remediation activities permitted by Sec. 50.59 or Sec. 53.1540,
``Updating licensing-basis information and determining the need for NRC
approval.'' Then, at least 2 years before expected license termination,
the licensee must submit an LTP for NRC review and approval. The
current decommissioning framework is predicated on the assumption that
power reactors will take a significant amount of time, potentially up
to 60 years, to terminate their license. However, the proposed
regulations would more readily and efficiently accommodate prompt
decommissioning of power reactors.
The NRC is proposing to amend the decommissioning licensing process
to allow licensees to submit the LTP within 2 years of permanent
cessation of operations (referred to as the ``upfront LTP''). In this
case, the licensee would not be required to submit the PSDAR. The goal
of this proposed change is to provide an option for increased
regulatory certainty, in the form of a license application and
approval, at the beginning of the decommissioning process for reactors
that choose to submit an upfront LTP. The NRC has received stakeholder
feedback from the power reactor industry that licensees perform the
majority of decommissioning activities ``at risk,'' without NRC review
and approval, before the submission of the LTP. At the time of NRC
review and approval of the licensing action (within 2 years of expected
license termination), the licensee has already removed the reactor
pressure vessel, cut and dismantled various reactor systems, and
remediated building surfaces and land areas. For many facilities, these
major decommissioning activities could potentially increase the amount
and variability of residual radioactivity that must be remediated and
surveyed before license termination. Consequently, this timing
challenge can reduce the efficiency of the decommissioning licensing
process, as licensees may need to implement additional corrective
measures to meet regulatory requirements after certain decommissioning
activities have been completed. For example, consider a hypothetical
site has completed its final status survey of an excavation and
subsequently backfilled it with offsite surface soil. If the NRC
determines that the final status survey is incomplete or inaccurate,
the licensee may not be able to selectively remove the clean backfill
in order to redo the survey. In another example, a hypothetical site
may perform radiation surveys of a building before demolition. If the
NRC identified a concern with the survey results, the licensee may need
to perform additional remediation activities near the building area,
which may have already been surveyed, as well. In such cases, NRC-
identified concerns of decommissioning activities performed ``at risk''
may delay license termination for certain facilities that had conducted
such decommissioning activities before meeting all regulatory
requirements.
Licensees that submit an upfront LTP within 2 years of permanent
cessation of operations would benefit from increased regulatory
certainty during the decommissioning process. Additionally, licensees
opting for this approach would not be required to submit a PSDAR, as
its three required components under Sec. 50.82(a)(4)(i) (e.g.,
schedule and description of the planned decommissioning activities,
discussion of environmental impacts, and site-specific decommissioning
cost estimate) are already captured within the eight elements of the
LTP under Sec. 50.82(a)(9)(ii), thereby streamlining the licensing
process and avoiding duplication of effort. When the PSDAR and LTP were
established in 1996 (61 FR 39301), the NRC noted that the PSDAR and LTP
could be combined into one submittal, which would have to be submitted
to NRC within 2 years of permanent shutdown. In this case, the
requirements for both the PSDAR and LTP would apply to the combined
document, including the requisite waiting period, public meeting, and
approval by amendment of the LTP. However, no power reactor licensee,
to date, has submitted a combined PSDAR and LTP. This proposed rule
would streamline the licensing process for facilities that are able to
prepare an LTP within 2 years of permanent cessation of operations.
This proposal is similar to the long-standing, proven
decommissioning licensing process of submitting a Decommissioning Plan
(DP) soon after permanent cessation of operations. This approach was
the standard decommissioning licensing framework for power reactor
licensees until the introduction of the PSDAR and LTP. The DP approach,
which is currently the required licensing mechanism for nonpower
production or utilization facilities and all nuclear materials
facilities, requires licensees to submit the DP for NRC review and
approval promptly after permanent cessation of operations. Under this
approach, the licensee could not perform dismantlement activities until
the approved DP amended the license and the NRC issued an associated
order permitting decommissioning activities in accordance with the
approved plan. However, the proposed revisions here differ from the DP
in that power reactor licensees would be able to perform
decommissioning activities consistent with Sec. 50.59 or Sec.
53.1540, as applicable, prior to the approval of the upfront LTP.
The NRC anticipates that future reactors licensed under 10 CFR
parts 50, 52, and 53 may be smaller, less technically complex, and may
operate for a shorter period of time than the current fleet of
traditional reactors, while maintaining adequate records and
decommissioning financial assurance. These facilities may plan for and
begin decommissioning activities promptly after permanent cessation of
operations, during operations, or even during application for an
operating license. For example, consider a hypothetical SMR, licensed
under 10 CFR part 52, which operates for 5 years in a remote location
to provide emergency power to a local community after a natural
disaster. Because the facility has a short operating history, has not
undergone reactor refueling, and has a small license footprint, the
licensee wishes to decommission promptly after cessation of operations.
The licensee plans to terminate the license within 2 years. Under the
current regulations, the licensee would either need to request an
exemption from the requirement to submit a PSDAR or submit both the
PSDAR and LTP after permanent shutdown. However, there would be no
increased safety benefit from submitting both a PSDAR and LTP within
this short timeframe. Under the proposed regulations, the licensee
could focus on developing and submitting a quality LTP to promptly
terminate the license in a safe manner. Accordingly, this proposed
licensing enhancement would significantly reduce administrative burden
on licensees who submit an upfront LTP.
The NRC foresees challenges for traditional large power reactor
licensees in developing an LTP within 2 years of permanent cessation of
operations. The long-standing PSDAR and LTP approach was developed
specifically for large power reactors to transition to decommissioning
with less regulatory burden and begin decommissioning activities with
sufficient funding promptly. However, upon entering decommissioning,
licensees must plan and safely perform large-scale dismantlement and
remediation activities that have not been attempted

[[Page 60728]]

during operations (e.g. removal of the reactor pressure vessel,
dismantlement of the containment building). Traditional large power
reactor licensees may need time to identify remaining dismantlement
activities and develop the historical site assessment, sufficient
characterization reports, and detailed plans for final radiation survey
to be included in the LTP. The NRC does not anticipate that the current
fleet of reactors would benefit from using the upfront LTP approach
because the licensee would need to develop a large volume of technical
information soon after operating a large facility for many decades.
Rather, the upfront LTP would be most beneficial for licensees who
expect to begin decommissioning promptly after permanent shutdown; have
less complex dismantlement, decommissioning planning, and radiation
survey activities; and can feasibly terminate the license in a
foreseeable timeframe. Nevertheless, the NRC is proposing to provide
the option to submit an upfront LTP to all power reactor facilities
licensed under 10 CFR parts 50, 52, and 53. Because both
decommissioning pathways are proven methods to meet NRC requirements
for license termination, the agency does not propose to limit the
upfront LTP option to a subset of licensees, but rather allow all power
reactor licensees to determine which approach best suits their
operational and planning needs.
Accordingly, this proposed rule has conforming changes to
Sec. Sec. 50.82(a)(5), 50.82(a)(7), 52.110(e), 52.110(g), 53.1070(e),
and 53.1070(g), which reference submittal of the PSDAR. This proposed
rule would update the requirements to reference the submittal required
before or within 2 years of permanent cessation of operations. The NRC
would retain the core content of the requirement, regardle

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