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



[[Page 86917]]

Vol. 89

Thursday,

No. 211

October 31, 2024

Part II





 Nuclear Regulatory Commission





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





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

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

[[Page 86918]]


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NUCLEAR REGULATORY COMMISSION

10 CFR Parts 1, 2, 10, 11, 19, 20, 21, 25, 26, 30, 40, 50, 51, 53, 
70, 72, 73, 74, 75, 95, 140, 150, 170, and 171

[NRC-2019-0062]
RIN 3150-AK31


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

AGENCY: Nuclear Regulatory Commission.

ACTION: Proposed rule.

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

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

ADDRESSES: You may submit comments by any of the following methods 
however, the NRC encourages electronic comment submission through the 
Federal rulemaking website:
     Federal Rulemaking website: Go to https://www.regulations.gov and search for Docket ID NRC-2019-0062. Address 
questions about NRC dockets to Helen Chang; telephone: 301-415-3228; 
email: [email protected]. For technical questions contact the 
individuals listed in the FOR FURTHER INFORMATION CONTACT section of 
this document.
     Email comments to: [email protected]. If you do 
not receive an automatic email reply confirming receipt, then contact 
us at 301-415-1677.
     Fax comments to: Secretary, U.S. Nuclear Regulatory 
Commission at 301-415-1101.
     Mail comments to: Secretary, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, ATTN: Rulemakings and 
Adjudications Staff.
     Hand deliver comments to: 11555 Rockville Pike, Rockville, 
Maryland 20852, between 7:30 a.m. and 4:15 p.m. eastern time, Federal 
workdays; telephone: 301-415-1677.
    You can read a plain language description of this proposed rule at 
https://www.regulations.gov/docket/NRC-2019-0062. For additional 
direction on obtaining information and submitting comments, see 
``Obtaining Information and Submitting Comments'' in the SUPPLEMENTARY 
INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Robert Beall, Office of Nuclear 
Material Safety and Safeguards, telephone: 301-415-3874; email: 
[email protected]; or Anders Gilbertson, Office of Nuclear Reactor 
Regulation, telephone: 301-415-1541; email: [email protected]. 
Both are staff of the U.S. NRC, Washington, DC 20555-0001.

SUPPLEMENTARY INFORMATION: 

Executive Summary

A. Need for the Regulatory Action

    On January 14, 2019, the President signed the Nuclear Energy 
Innovation and Modernization Act (NEIMA) into law (Pub. L. 115-439). 
NEIMA section 103(a)(4) directs the NRC to ``complete a rulemaking to 
establish a technology-inclusive, regulatory framework for optional use 
by commercial advanced nuclear reactor applicants for new reactor 
license applications.'' NEIMA defines a ``technology-inclusive 
regulatory framework'' as one that is ``developed using methods of 
evaluation that are flexible and practicable for application to a 
variety of reactor technologies, including, where appropriate, the use 
of risk-informed and performance-based techniques.'' NEIMA, as further 
amended by the Accelerating Deployment of Versatile, Advanced Nuclear 
for Clean Energy Act of 2024 (ADVANCE Act), defines the term ``advanced 
nuclear reactor'' as ``a nuclear fission reactor or fusion machine, 
including a prototype plant (as defined in sections 50.2 and 52.1 of 
title 10, Code of Federal Regulations (as in effect on the date of 
enactment of [NEIMA])), with significant improvements compared to 
commercial nuclear reactors under construction as of the date of 
enactment of [NEIMA].''
    The NRC initially considered establishing the scope of proposed 
part 53, ``Risk-Informed, Technology-Inclusive Regulatory Framework for 
Commercial Nuclear Plants,'' of title 10 of the Code of Federal 
Regulations (10 CFR) as being for ``advanced nuclear plants'' 
consisting of one or more ``advanced nuclear reactors'' as defined in 
NEIMA. Based on public discussions on the use of the term, the NRC 
determined that the NEIMA definition, although broad, did not define 
``significant improvements'' with enough specificity to implement in 
NRC regulations. Additionally, a number of stakeholders suggested that 
the descriptor, ``advanced,'' implied enhanced safety, while the NEIMA 
definition includes ``significant improvements'' in areas other than 
safety enhancements. In response to this feedback, and to be technology 
inclusive, the NRC determined that the broader term ``commercial 
nuclear plant'' would be preferable.
    The current application and licensing requirements in 10 CFR part 
50, ``Domestic Licensing of Production and Utilization Facilities,'' 
and 10 CFR part 52, ``Licenses, Certifications, and Approvals for 
Nuclear Power Plants,'' were primarily developed to address license 
requests concerning water-cooled reactors, and to address operational 
requirements for those types of reactors. This proposed rule responds 
to NEIMA by creating an alternative regulatory framework for licensing 
future commercial nuclear plants. The new alternative requirements and 
implementing guidance would adopt technology-inclusive approaches and 
use risk-informed and performance-based techniques to ensure an 
equivalent level of safety to that of operating commercial nuclear 
plants while providing flexibility for licensing and regulating a 
variety of technologies and designs for commercial nuclear reactors.

B. Major Provisions

    Major provisions of this proposed rule, supported by accompanying 
guidance, include the following:
     A new alternative technology-inclusive, risk-informed, 
performance-based framework that includes requirements for licensing 
and regulating nuclear plants during the various stages of their life 
cycles.
     A new alternative technology-inclusive, risk-informed, and 
performance-based framework in 10 CFR part 26, ``Fitness for Duty 
Programs,'' developed from existing requirements in subpart K, ``FFD 
Programs for Construction,'' of part 26.
     A new alternative technology-inclusive and performance-
based security framework in 10 CFR part 73, ``Physical Protection of 
Plants and Materials,'' that includes requirements for protection of 
licensed activities at commercial nuclear plants.

C. Costs and Benefits

    The NRC prepared a draft regulatory analysis to determine the 
expected quantitative costs and benefits of this proposed rule and 
associated guidance as well as qualitative factors to be considered in 
the NRC's rulemaking decision. The conclusion from the

[[Page 86919]]

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

Table of Contents

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

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2019-0062 when contacting the NRC 
about the availability of information for this action. You may obtain 
publicly available information related to this action by any of the 
following methods:
     Federal Rulemaking Website: Go to https://www.regulations.gov and search for Docket ID NRC-2019-0062.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly available documents online in the 
ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``Begin Web-based ADAMS 
Search.'' For problems with ADAMS, please contact the NRC's Public 
Document Room (PDR) reference staff at 1-800-397-4209, at 301-415-4737, 
or by email to [email protected]. For the convenience of the reader, 
instructions about obtaining materials referenced in this document are 
provided in the ``Availability of Documents'' section.
     NRC's PDR: The PDR, where you may examine and order copies 
of publicly available documents, is open by appointment. To make an 
appointment to visit the PDR, please send an email to 
[email protected] or call 1-800-397-4209 or 301-415-4737, between 8 
a.m. and 4 p.m. eastern time, Monday through Friday, except Federal 
holidays.

B. Submitting Comments

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

II. Background

A. NRC Advanced Reactor Readiness

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

[[Page 86920]]

provided the Commission's policy regarding the review of, and desired 
characteristics associated with, advanced reactors. The NRC updated 
this statement in the ``Policy Statement on the Regulation of Advanced 
Reactors,'' dated October 14, 2008 (Advanced Reactor Policy Statement).
    The agency has undertaken many activities related to advanced 
reactors, including issuing an advance notice of proposed rulemaking 
titled, ``Approaches to Risk-Informed and Performance-Based 
Requirements for Nuclear Power Reactors,'' dated May 4, 2006 (71 FR 
26267). These efforts were often done in parallel, and sometimes 
interwoven, with the NRC's efforts to improve risk-informed and 
performance-based approaches within the agency (e.g., the Commission's 
policy statement, ``Use of Probabilistic Risk Assessment Methods in 
Nuclear Regulatory Activities,'' dated August 16, 1995 (PRA Policy 
Statement)).
    In 2016, the NRC issued ``NRC Vision and Strategy: Safely Achieving 
Effective and Efficient Non-Light-Water Mission Readiness'' (Advanced 
Reactor Vision and Strategy Document), in response to increasing 
interest in advanced reactor designs. The NRC considered the Department 
of Energy's (DOE's) advanced reactor deployment goals in developing the 
Advanced Reactor Vision and Strategy Document. Since publication of the 
document, the NRC continues to manage its activities to support the 
DOE's deployment goals. The Advanced Reactor Vision and Strategy 
Document identified initiating and developing a new risk-informed and 
performance-based regulatory framework as a possible long-term goal. 
However, the NRC staff's initial efforts were focused on resolving 
policy issues and developing guidance for licensing non-LWR 
technologies under the existing regulatory frameworks (parts 50 and 
52). The NRC staff issues annual Commission papers on the status and 
progress of the NRC staff's activities related to advanced reactors 
(e.g., SECY-24-0020, ``Advanced Reactor Program Status,'' dated 
February 27, 2024). These Commission papers provide status updates for 
advanced reactor activities undertaken both prior to and after 
initiation of this rulemaking.
    In 2017, the NRC staff prioritized activities to support the 
development of technology-inclusive, risk-informed, and performance-
based licensing approaches that could be implemented under the existing 
regulatory framework in parts 50 and 52. One key element of these 
efforts was the Licensing Modernization Project (LMP), a cost-shared 
initiative led by nuclear utilities and supported by DOE. The LMP is a 
technology-inclusive, risk-informed, and performance-based methodology 
developed for non-LWR designs. The LMP provides a systematic and 
reproducible process for licensing-basis event (LBE) selection and 
evaluation; classification of structures, systems, and components 
(SSCs); and assessment of defense in depth. The LMP refined the DOE's 
Next Generation Nuclear Plant Program methodologies to reflect 
interactions with the NRC, to address feedback from industry, and to 
broaden the scope of the approach to ensure applicability to various 
non-LWR technologies. The LMP activities led to the publication and 
submittal of Nuclear Energy Institute (NEI) 18-04, Revision 1, ``Risk-
Informed Performance-Based Technology Inclusive Guidance for Non-Light 
Water Reactor Licensing Basis Development,'' issued August 2019. The 
document indicates that controlling the frequencies and potential 
consequences of a wide spectrum of events is the primary focus of the 
LMP approach.
    The NRC endorsed the principles and methodology in NEI 18-04, with 
clarifications, in RG 1.233, ``Guidance for a Technology-Inclusive, 
Risk-Informed, and Performance-Based Methodology to Inform the 
Licensing Basis and Content of Applications for Licenses, 
Certifications, and Approvals for Non-Light-Water Reactors.'' The NRC 
staff sought Commission approval of the use of LMP and NEI-18-04 in 
SECY-19-0117, ``Technology-Inclusive, Risk-Informed, and Performance-
Based Methodology to Inform the Licensing Basis and Content of 
Applications for Licenses, Certifications, and Approvals for Non-Light-
Water Reactors,'' dated December 2, 2019. In that paper, the staff 
described the relationship between the LMP and NEI-18-04 and previous 
relevant Commission decisions, including those described in SECY-93-
092, ``Issues Pertaining to the Advanced Reactor (PRISM, MHTGR, and 
PIUS) and CANDU 3 Designs and their Relationship to Current Regulatory 
Requirements,'' dated April 8, 1993. The Commission approved the use of 
the LMP methodology and NEI-18-04 as a reasonable approach for 
establishing key parts of the licensing basis and content of 
applications for licenses, certifications, and approvals for non-LWRs 
in Staff Requirements Memorandum (SRM) SRM-SECY-19-0117, dated May 26, 
2020. Although the LMP approach is technology- inclusive, the industry 
and NRC staff initially focused the LMP's applicability on non-LWRs, 
both for efficiency and to support near-term non-LWR applications under 
the existing regulatory framework, such as the Advanced Reactor 
Demonstration Projects supported by DOE.
    As stated in the part 53 rulemaking plan, SECY-20-0032, the NRC 
staff developed part 53 by building upon recent and ongoing activities 
such as the LMP approach described in SECY-19-0117. Such an approach 
supports implementing the NEIMA requirement to use, where appropriate, 
risk-informed and performance-based techniques, and it also capitalizes 
on previous initiatives by the industry, DOE, and the NRC, including 
the LMP. This approach highlights the role of PRA in risk-informed and 
performance-based approaches to identifying enhanced safety margins 
that can be used to justify operational flexibilities. The proposed 
framework is largely based on the methodology described in SECY-19-0117 
and includes a prominent role for PRA.
    As discussed in section II.B, ``Stakeholder Views on Part 53 
Preliminary Proposed Rule Language,'' of this document, the NRC 
conducted extensive public outreach on early versions of the proposed 
rule text. Early versions of the draft proposed rule included two 
alternative regulatory frameworks. One framework (called ``Framework 
A'') offered a licensing approach centered largely on risk analysis and 
the other framework (called ``Framework B'') largely replicated the 
existing licensing approach in parts 50 and 52 but modified it to be 
technology neutral. In its SRM to SECY-23-0021, ``Proposed Rule: Risk-
Informed, Technology-Inclusive Regulatory Framework for Advanced 
Reactors (RIN 3150-AK31),'' the Commission disapproved the inclusion of 
Framework B in this proposed rule and directed the staff to provide 
them within one year an options paper for possible future use of the 
Framework B methodology.

B. Stakeholder Views on Part 53 Preliminary Proposed Rule Language

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

[[Page 86921]]

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

III. Discussion

A. Objective and Applicability

    The NRC is proposing to add a new, alternative part to its 
regulations that would set out a risk-informed, technology-inclusive 
framework for the licensing and regulation of commercial nuclear 
plants. This new approach would achieve the following: (1) continue to 
provide reasonable assurance of adequate protection of public health 
and safety and the common defense and security; (2) promote regulatory 
stability, predictability, and clarity; (3) reduce requests for 
exemptions from the current requirements in parts 50 and 52; (4) 
establish new requirements to address non-LWR technologies; (5) 
recognize technological advancements in reactor design; and (6) credit 
the possible response of some designs of commercial nuclear plants to 
postulated accidents, including slower transient response times and 
relatively small and slow release of fission products. This proposed 
rule would add 10 CFR part 53; subpart M, ``Fitness for Duty Programs 
for Facilities Licensed Under 10 CFR Part 53,'' to Part 26; Sec.  
73.100, ``Technology-inclusive requirements for physical protection of 
licensed activities at commercial nuclear plants against radiological 
sabotage,'' Sec.  73.110, ``Technology-inclusive requirements for 
protection of digital computer and communication systems and 
networks,'' and Sec.  73.120, ``Access authorization program for 
commercial nuclear plants,'' as well as make conforming changes 
throughout 10 CFR chapter I, ``Nuclear Regulatory Commission.''

B. Need for Changes to the Existing Regulatory Framework

    The NRC has long recognized that the licensing and regulation of a 
variety of nuclear reactor technologies would present challenges 
because the existing regulatory framework has evolved primarily to 
address the LWR designs that compose the current operating fleet 
(widely referred to as Generation II reactors). The NRC has had many 
interactions with designers of various reactor technologies under 
development, sometimes collectively referred to as advanced reactors 
(widely referred to as Generation III/III+ (i.e., evolutionary light-
water) and Generation IV (i.e., non-light-water) reactors). The 
interactions have informed the development of policies and guidance to 
support the potential licensing of new and different types of reactor 
facilities, some of which may not utilize LWR designs. The NRC issued 
its Advanced Reactor Policy Statement to provide all interested 
parties, including the public, with the Commission's views concerning 
the desired characteristics of advanced reactor designs. The NRC 
further described its early efforts to establish a technology-inclusive 
approach to the regulation of nuclear reactors in the advance notice of 
proposed rulemaking published in 2006. The NRC acknowledged in its 
``Report to Congress: Advanced Reactor Licensing,'' issued August 2012, 
that while the safety philosophy inherent in the current regulations 
applies to all reactor technologies, the specific and prescriptive 
aspects of those regulations clearly focus on the current fleet of LWR 
facilities.
    Congress similarly recognized the potential benefits of developing 
a regulatory infrastructure to support the development and 
commercialization of advanced nuclear reactors. Consequently, Congress 
passed NEIMA in late 2018, and the President signed it into law in 
January 2019. NEIMA directed the NRC to undertake a rulemaking to 
establish a technology-inclusive regulatory framework for optional use 
by applicants for new commercial advanced nuclear reactor licenses. In 
addition, on July 9, 2024, the President signed into law the 
Accelerating Deployment of Versatile, Advanced Nuclear for Clean Energy 
Act of 2024, also referred to as the ADVANCE Act. The NRC is evaluating 
its plans for implementing the ADVANCE Act, including how its 
regulations, as well as the proposed part 53 or future revisions to it, 
could be used to address provisions in the ADVANCE Act. The ADVANCE Act 
contains provisions on a variety of nuclear-related topics, such as 
micro reactors, nuclear reactor license application reviews, and 
nuclear fuel. In Section VI, ``Specific Requests for Comments,'' the 
NRC is requesting public input on how part 53 could be revised to 
better enable its potential use to implement the ADVANCE Act.
    The requirements in part 53 would support a wide variety of 
potential commercial nuclear reactor technologies. As noted in this 
discussion, the current regulatory framework in parts 50 and 52 evolved 
in the context of the current operating reactor fleet dominated by LWRs 
and as a result includes provisions specific to LWR technologies. While 
the NRC can license other reactor technologies under the current 
framework by using existing regulatory flexibilities and the exemption 
process, there is significant interest in developing a regulatory 
framework that is flexible enough to accommodate multiple technologies 
and robust enough to ensure a level of safety equivalent to parts 50 
and 52, consistent with the Commission's Advanced Reactor Policy 
Statement. The Commission reiterated its safety

[[Page 86922]]

expectations for new reactors in the SRM for SECY-10-0121, ``Modifying 
the Risk-Informed Regulatory Guidance for New Reactors,'' dated March 
2, 2011:

    Because new plant designs incorporate operating experience from 
current generation reactors, severe accident research, and risk 
insights from design probabilistic risk assessments, the Commission 
expects that the advanced technologies incorporated in new reactors 
will result in enhanced margins of safety. However, the Commission 
continues to expect (consistent with the 2008 Advanced Reactor 
Policy Statement), as a minimum, at least the same degree of 
protection of the public and the environment that is required for 
current-generation light-water reactors. New reactors with these 
enhanced margins and safety features should have greater operational 
flexibility than current reactors.

    However, developing a regulatory framework that can accommodate a 
wide range of technologies while maintaining an acceptable level of 
safety presents significant regulatory challenges. The existing 
regulations have been developed over the course of decades and reflect 
changes to address events discovered through operating experience. In 
contrast, part 53 is being developed to accommodate technologies that, 
in some cases, lack significant operating experience. To address these 
challenges, the NRC drew on well-developed approaches to licensing to 
produce a technology-neutral and robust regulatory framework. The 
proposed regulatory framework would use PRAs to assess risks, help 
establish technical requirements, and manage operations. The framework 
builds on the LMP, which is a technology-inclusive approach to 
licensing that leverages insights from a detailed PRA to provide 
applicants with significant design and operation flexibilities.

C. 10 CFR Part 53: Framework

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

IV. Part 53: Framework

Subpart A--General Provisions

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

[[Page 86923]]

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

[[Page 86924]]

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

Subpart B--Technology-Inclusive Safety Requirements

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

[[Page 86925]]

minimum. The requirements throughout part 53 that support demonstrating 
compliance with Sec.  53.220 would be similar to current regulations 
that both contribute to assuring adequate protection of public health 
and safety and are desirable to promote the common defense and security 
or to protect health or to minimize danger to life or property under 
Section 161 of the Act.
    Consistent with historical practice, Sections 182 and 161 of the 
Act are cited as authorizing legislation within this proposed rule. 
However, specific language from the Act would not be incorporated into 
the safety objectives or safety criteria in part 53. This is because, 
again consistent with historical practice, the NRC would not be 
defining ``adequate protection'' through the individual safety 
requirements in part 53. Rather, part 53 would enable the NRC to make 
its required findings under the Act by providing sufficient performance 
standards, safety criteria, and related requirements on how applicants 
must demonstrate compliance with subpart B and other subparts.
    Section 53.210 would provide safety criteria for DBAs that would be 
required to be identified under Sec.  53.240 and analyzed under Sec.  
53.450(f) in subpart C of part 53. Subsequent sections in part 53 would 
require that the SSCs relied upon to demonstrate compliance with the 
criteria in Sec.  53.210 be classified as SR. The use of SR SSCs and 
the 25 rem reference values for potential radiological consequences 
would align with traditional deterministic approaches for LWRs from 
Sec. Sec.  50.34, 52.79, and 100.11 for evaluating the effectiveness of 
plant design features with respect to postulated reactor accidents. A 
footnote similar to that included in Sec.  50.34(a)(1)(ii)(D)(1) and 
Sec.  52.79(a)(1)(vi)(A) would be included in Sec.  53.210 to explain 
that the use of the 25 rem value would not be intended to imply that 
this number constitutes an acceptable limit for an emergency dose to 
the public under accident conditions. Rather, this dose value has been 
set forth in this proposed section as a reference value that would be 
used in the evaluation of plant design features with respect to DBAs to 
verify that the proposed designs would provide assurance of low risk of 
public exposure to radiation in the event of an accident. The inclusion 
of the safety criteria for DBAs in subpart B would provide a logical 
structure supporting the identification and treatment of SR SSCs and 
establishing the corresponding functional design criteria for those 
SSCs.
    Section 53.220 would provide safety criteria for LBEs other than 
DBAs that would be required to be identified under Sec.  53.240 and 
analyzed under Sec.  53.450(e) in subpart C. Whereas Sec.  53.210 and 
the related requirements for SR SSCs would provide that a defined 
success path exists for DBAs, the safety criteria for LBEs other than 
DBAs would establish the connections between SSC design, human actions, 
and programmatic controls and a broader set of potential internal and 
external hazards. These safety criteria would also address defense-in-
depth matters such as a balanced consideration of prevention and 
mitigation.
    The safety criterion in Sec.  53.220(b) would include a requirement 
to use a comprehensive risk metric or set of metrics and associated 
risk performance objectives against which calculated values of the risk 
metrics are compared. The comprehensive risk metrics or set of metrics 
and associated risk performance objectives would support a performance-
based approach to developing an appropriate combination of design 
features and programmatic controls to prevent or mitigate LBEs other 
than DBAs. The applicant must propose the comprehensive risk metric or 
set of metrics and associated risk performance objectives, and the 
comprehensive risk metric or set of metrics and associated risk 
performance objectives must provide an appropriate level of safety. 
Comprehensive risk metrics should consist of a proposed plant risk 
metric or set of proposed risk metrics that approximate the total, 
overall risk from the facility and that address the range of possible 
plant configurations and associated internal and external hazards to 
the extent practicable. The associated risk performance objectives are 
preestablished, indicative values of the comprehensive risk metrics 
that are used as part of risk-informed decision-making. The methodology 
for developing and using proposed comprehensive risk metrics and 
associated risk performance objectives is defined by the proposed 
requirements for analyses in Sec.  53.450. Therefore, the application 
must include a description of that methodology and, among other things, 
should explain the initial conditions, boundary conditions, and key 
assumptions used to develop and calculate the risk metrics. Screening 
tools and bounding or simplified methods may be used for any mode or 
hazard, provided that the applicant provides an acceptable technical 
basis. As with all risk-informed methodologies, treatment of 
uncertainties must be addressed.
    The risk performance objectives established under this methodology 
are likely to involve assessing and averaging the risks over a period 
of time (e.g., plant year) and would not constitute a real-time 
requirement that must be continuously demonstrated by the licensee. The 
use of a comprehensive risk metric or set of risk metrics and risk 
performance objectives that reflect an average risk to establish 
performance goals for SR and NSRSS SSCs is consistent with current 
practices that use other risk assessment techniques to address short-
term plant configurations during plant maintenance activities.
    It is worth noting that the evaluation of plant risks, as 
represented by a comparison of analysis results to acceptable risk 
performance objectives for comprehensive risk metrics, would be one of 
several performance standards used in subpart B. The proposed use of 
multiple performance standards, including deterministic criteria and 
defense-in-depth measures, reflects an integrated decision-making 
process similar to that described in RG 1.174, ``An Approach for Using 
Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-
Specific Changes to the Licensing Basis,'' Revision 3. The NRC's 
approval of using a comprehensive risk metric or set of metrics with 
associated risk performance objectives is not, by itself, an indicator 
of adequate protection. Rather, the comparison of comprehensive risk 
metrics to associated risk performance objectives that are acceptable 
to the NRC is part of a suite of regulatory requirements that, when 
considered holistically, form the basis for the NRC's decision-making. 
This is analogous to the approach used for plants licensed under part 
50 and part 52, where no single regulatory requirement governs whether 
a plant is ``safe enough.''
    The RG 1.233, ``Guidance for a Technology-Inclusive, Risk-Informed, 
and Performance-Based Methodology to Inform the Licensing Basis and 
Content of Applications for Licenses, Certifications, and Approvals for 
Non-Light-Water Reactors,'' describes an example of an acceptable 
approach for identifying and analyzing LBEs under part 50 and part 52, 
including the use of the quantitative health objectives (QHOs) stated 
in the NRC's policy statement, ``Safety Goals for Nuclear Power Plant 
Operation,'' dated August 4, 1986 (51 FR 28044), as corrected and 
republished August 21, 1986 (51 FR 30028) (Safety Goals Policy 
Statement), as acceptable performance objectives for

[[Page 86926]]

comprehensive risk metrics. The use of comprehensive risk metrics, such 
as the individual early fatality risk (IEFR) and the individual latent 
cancer fatality risk (ILCFR), and associated risk performance 
objectives, such as the QHOs, from the Safety Goals Policy Statement, 
could form the basis for one approach to meet Sec.  53.220(b). The 
requirement for comprehensive risk metrics, in combination with the 
other proposed requirements in subparts B and C, would bring the 
approach endorsed in RG 1.233 for parts 50 and 52 into part 53. 
Additionally, the use of comprehensive risk metrics and associated risk 
performance objectives would provide a logical performance objective to 
support the risk management approaches in the various subparts 
comprising proposed part 53.
    The Commission stated in the introduction of the Safety Goals 
Policy Statement that improvements to then-current regulatory practices 
could lead to a more coherent and consistent regulation of nuclear 
power plants, a more predictable regulatory process, a better public 
understanding of the regulatory criteria that the NRC applies, and 
public confidence in the safety of operating plants. Accordingly, the 
Commission announced the safety goals with a focus on the risks to the 
public from nuclear power plant operation. Following the issuance of 
the Safety Goals Policy Statement, the NRC has used the comprehensive 
risk metrics and performance objectives provided in the safety goals 
within the criteria for many decisions involving safety judgments 
during the licensing and regulation of operating reactors and proposed 
nuclear reactor designs. Consistent with NUREG-0880, the proposed 
comprehensive risk metrics and associated risk performance objectives 
required under Sec.  53.220(b) could be expressed in terms of a 
biologically average individual in terms of age and other risk factors. 
Although some comprehensive risk objectives such as the IEFR and ILCFR 
are defined in terms of fatality risks, the Commission continues to 
make clear that no death attributable to nuclear power plant operation 
will ever be ``acceptable'' in the sense that the Commission would 
regard it as a routine or permissible event. Comprehensive risk metrics 
and associated risk performance objectives as used in this proposed 
rule would establish acceptable risks, not acceptable deaths.
    Applicants under the proposed part 53 may choose to develop and 
seek NRC approval of comprehensive risk metrics or sets of risk metrics 
and associated risk performance objectives beyond those discussed 
above, including the use of surrogate measures for use in specific 
analyses to satisfy the proposed requirements in Sec.  53.220(b). Such 
surrogate measures for comprehensive risk metrics and associated risk 
performance objectives could be used in a manner similar to the use of 
core damage frequency and conditional containment failure probability 
for LWRs within the safety goal evaluation process in NUREG/BR-0058, 
``Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory 
Commission,'' and other assessments of LWRs using the NRC's safety 
goals. The NRC would, as appropriate, review novel approaches for 
comprehensive metrics and associated risk performance goals proposed by 
applicants, industry organizations, or standard development 
organizations and would engage stakeholders during the development of 
the related regulatory guidance or specific licensing actions.
    Section 53.230 would require safety functions needed to ensure that 
the safety criteria under Sec. Sec.  53.210 and 53.220 can be met if an 
assumed LBE were to occur at a commercial nuclear plant. Section 53.230 
would specify that limiting the release of radioactive materials from 
the facility is the primary safety function, and therefore, limiting 
potential offsite consequences (i.e., dose to a hypothetical 
individual) would be used as the primary performance metric throughout 
part 53. The additional or subsidiary safety functions needed to limit 
the release of radionuclides may include, without limitation, 
controlling processes related to reactivity, heat generation, heat 
removal, and chemical interactions. This proposed rule provides 
flexibility to applicants and licensees in identifying, implementing, 
and maintaining the safety functions supporting retention of 
radionuclides for commercial nuclear plants of varying sizes and 
technologies.
    Proposed Sec.  53.240 would require applicants to identify and 
address LBEs. LBEs are unplanned events, resulting from both internal 
and external hazards, that are used in the design and analyses required 
under part 53 for licensing commercial nuclear plants. This ensures 
estimates of offsite consequences from analyses performed under 
proposed Sec.  53.450 are below the safety criteria identified under 
proposed Sec. Sec.  53.210 and 53.220 and that SSCs, personnel, and 
programs address the safety functions from proposed Sec.  53.230. 
Including a high-level performance requirement related to the 
identification and analysis of LBEs in subpart B would reflect the 
historical and continuing importance of evaluating unplanned events as 
part of the licensing of commercial nuclear plants. Proposed Sec.  
53.240 would require identification and analysis of LBEs under Sec.  
53.450, which would require a PRA. Examples of acceptable methods of 
using PRAs to identify and assess LBEs would be the methodology in RG 
1.233, as discussed in Draft Regulatory Guide (DG)-1413, ``Technology-
Inclusive Identification of Licensing Events for Commercial Nuclear 
Plants.''
    Section 53.250 would establish defense-in-depth requirements based 
on the longstanding philosophy of providing defense in depth to address 
uncertainties about the design, operation, and performance of 
commercial nuclear plants. For example, parts 50 and 52 address defense 
in depth through layered prescriptive technical requirements (e.g., 
fuel performance, cladding integrity, reactor coolant system integrity, 
containment performance) for LWRs. In contrast, the flexibility 
afforded to applicants in how they propose to demonstrate compliance 
with the high-level safety criteria within part 53 would necessitate 
this specific requirement to ensure defense in depth is provided. The 
requirements in this section would state that no single engineered 
design feature, human action, or programmatic control, no matter how 
robust, should be exclusively relied upon to address LBEs other than 
DBAs. The phrase ``engineered design feature'' would not preclude the 
possible crediting of inherent characteristics within the design and 
analysis for commercial nuclear reactors. While defense in depth would 
only be assessed for LBEs other than DBAs, the need to ensure dedicated 
success paths for DBAs would contribute to the overall defense in depth 
for each commercial nuclear plant under part 53.
    Section 53.260 would govern normal operations and would establish a 
level of safety based on current requirements in 10 CFR part 20, 
``Standards for Protection Against Radiation,'' which limits doses to 
members of the public and dose rates in unrestricted areas.
    Section 53.270 would provide for the protection of plant workers 
and would establish a level of safety based on current requirements in 
10 CFR part 20 which limits occupational dose.

Subpart C--Design and Analysis Requirements

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

[[Page 86927]]

subpart B can be satisfied. The sections within subpart C would reflect 
the overall hierarchy throughout part 53, which would cover: (1) plant-
level safety criteria (Sec. Sec.  53.210, 53.220, and 53.470); (2) 
safety functions (Sec.  53.230) needed to demonstrate compliance with 
the safety criteria; (3) design features (Sec.  53.400), human actions, 
and programmatic controls needed to fulfill the safety functions; and 
(4) functional design criteria (Sec. Sec.  53.410 and 53.420) that must 
be defined for each design feature relied on to demonstrate the safety 
criteria (Sec. Sec.  53.210, 53.220, and 53.470) are met. Subpart C 
would also contribute to the logic and structure of part 53 by 
distinguishing between SR SSCs and NSRSS SSCs and licensee-controlled 
programs that address LBEs other than DBAs. Specifically, SR SSCs, 
human actions, and programmatic controls needed to protect against DBAs 
are used to satisfy the safety criteria in Sec.  53.210. Non-safety-
related but safety-significant SSCs, human actions, and licensee-
controlled programs that address LBEs other than DBAs generally 
contribute to the appropriate measures considering potential risks to 
public health and safety.
    Section 53.400 would establish a requirement that design features 
be provided for each commercial nuclear plant to satisfy the safety 
criteria and fulfill safety functions from proposed subpart B during 
LBEs. Other sections in subpart C would, in turn, further address the 
necessary capabilities and reliabilities for SSCs by establishing 
functional design criteria, fulfilling design requirements, performing 
analyses of LBEs, performing other supporting analyses, and 
categorizing SSCs based on their roles in preventing or mitigating 
LBEs.
    Section 53.410 would require that functional design criteria be 
defined for design features relied upon to demonstrate that the 
consequences from DBAs would be below the criteria in Sec.  53.210 
through analyses performed under Sec.  53.450(f), which includes 
insights from both PRAs and deterministic analyses. Other sections 
within part 53 would establish appropriate controls on these design 
features (e.g., safety classification, protection from external 
hazards, quality assurance, and TS) to ensure the functional design 
criteria are satisfied. The performance requirements for the SSCs 
needed to address DBAs and the corresponding human actions and 
programmatic controls would contribute to ensuring that a commercial 
nuclear plant licensed under part 53 would meet the safety criteria in 
Sec.  53.210.
    Section 53.415 would require that SR SSCs be protected against or 
designed to withstand the effects of natural phenomena (e.g., 
earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches) and 
constructed hazards (e.g., from dams, transportation routes, and 
military or industrial facilities). Specifically, Sec.  53.415 would 
require that SR SSCs remain capable of performing the safety functions 
stated in Sec.  53.230 for which they are credited up to the design-
basis external hazard levels as determined under Sec.  53.510. As used 
in Sec.  53.415 and subpart D of part 53, a hazard level would refer to 
such things as the magnitude and recurrence rate of an earthquake and 
the resultant ground motions, the height of a flood, the force of 
hurricane winds, or the concentrations of chemicals resulting from a 
release from a nearby facility. These requirements would support either 
traditional deterministic approaches for determining and protecting 
against external hazards or probabilistic approaches that are being 
developed for seismic and some other external hazards.
    Section 53.420 would require that functional design criteria be 
defined for design features that play a significant role in 
demonstrating that the safety criteria for LBEs other than DBAs are 
satisfied. The analyses required for this demonstration would be 
described in proposed Sec.  53.450(e), which would require that those 
events be identified and assessed using a PRA methodology in 
combination with other generally accepted approaches for systematically 
evaluating engineered systems. The SSCs determined to be safety 
significant (i.e., either SR or NSRSS) would have associated special 
treatment requirements as specified in Sec.  53.460. Special treatment 
would be defined in subpart A of part 53 and generally refers to 
measures (e.g., quality assurance, testing, monitoring) taken beyond 
the procurement and installation of commercial grade products to 
provide confidence that the SSC will comply with the applicable 
functional design criteria. The inclusion of a systematic approach to 
identifying the functional design criteria for SSCs and tailoring the 
special treatments to specific LBEs and safety functions is an 
important contributor to satisfy the proposed safety criteria in 
subpart B. Therefore, designers and licensees for commercial nuclear 
plants would be provided flexibility on how LBEs other than DBAs are 
either prevented or mitigated and how the calculated comprehensive 
plant risks satisfy the safety criterion established under Sec.  
53.220(b).
    Section 53.425 would establish requirements for design features and 
related functional design criteria limiting doses to members of the 
public during normal operations to satisfy the criteria in part 20. 
Section 53.430 would provide similar requirements for design features 
and related functional design criteria for protection of plant workers 
to meet the safety criteria in part 20. Similar to existing 
regulations, the NRC considers that licensees would generally comply 
with the requirements of part 20 to keep doses as low as reasonably 
achievable by meeting a design objective of keeping doses to the public 
from routine plant effluents less than 10 millirem per year. This goal 
is similar to that provided by appendix I to part 50 and would assist 
designers, applicants, and licensees in performing the evaluations of 
possible reductions in public dose from routine effluents when 
considering costs and other factors. As emphasized in existing 
regulations in part 50, the design objective of keeping doses to the 
public from routine plant effluents less than 10 millirem per year 
should not be construed as a radiation protection standard. The NRC 
anticipates that future guidance will continue to reflect this 
performance goal.
    The proposed requirements in Sec. Sec.  53.425 and 53.430 for 
design features and functional design criteria to support radiation 
protection activities have parallels in existing regulations such as 
Sec.  50.34(a) and (b)(3), which require in part that the means be 
provided for meeting the requirements of part 20 and General Design 
Criterion 60, 61, 63, and 64 in appendix A to part 50, which provide 
radiation protection related design criteria.
    Section 53.440 would address various design requirements that 
warrant specific mention to ensure that the design features required by 
Sec.  53.400 comply with the functional design criteria required by 
Sec. Sec.  53.410 and 53.420. These requirements would be met through 
design practices, consideration of testing and operating experience, 
and various assessments of LBEs and other potential challenges to 
commercial nuclear plants. Discussions of some of the key design 
requirements included in this section follow.
     Sec.  53.440(a): An essential element to ensuring a 
proposed design can comply with the performance criteria in proposed 
part 53 would be that the abilities of design features to fulfill their 
safety functions are demonstrated by a combination of analyses, test 
programs, prototype testing, and operating experience. This requirement 
closely aligns with the language in Sec.  50.43(e)

[[Page 86928]]

and is proposed in part 53 as the same foundational requirement. In 
addition, the proposed Sec.  53.440(a) would require the design 
processes for SSCs under this section to include administrative 
procedures for evaluating operating, design, and construction 
experience for considering applicable important industry experiences in 
the design of those SSCs. This proposed requirement corresponds to the 
existing requirement under Sec.  50.34(f)(3)(i) that was developed in 
response to the 1979 accident at Three Mile Island Nuclear Generating 
Station.
     Sec.  53.440(b): The design and licensing of commercial 
nuclear plants should use generally accepted consensus codes and 
standards. Such codes and standards ensure sufficient testing and 
qualification of materials and equipment and provide defined processes, 
specifications, and acceptance criteria for use by designers and 
suppliers. The NRC would indicate acceptance of consensus codes and 
standards used in the design and licensing of a specific commercial 
nuclear plant either through the NRC's generic endorsement of a code or 
standard (i.e., through regulatory guidance), including any limitations 
or conditions, that can be referenced within an application, or through 
the review of a referenced code or standard as part of the review of a 
specific application.
     Sec.  53.440(c): The design requirements in subpart C 
would require the materials used for SR and NSRSS SSCs to be qualified 
for their service conditions over the design life of the SSC.
     Sec.  53.440(d): The requirements in Sec.  53.440 would 
include the need to consider possible degradation mechanisms for 
materials and equipment to inform both the design process and the 
development of integrity assessment programs to be executed during 
plant operations in accordance with subpart F of part 53. The inclusion 
of requirements related to designing and monitoring for possible 
degradation mechanisms reflects important lessons learned from the 
history of LWRs as well as operating experience with structures and 
systems in countless other engineering endeavors.
     Sec.  53.440(e) and (f): The design requirements in 
subpart C would state specific design requirements similar to existing 
requirements in parts 50, 52, and 73 for protections against fires and 
explosions and consideration of safety and security together in the 
design process.
     Sec.  53.440(g) and (h): Specific design requirements are 
proposed to ensure that commercial nuclear reactors under part 53 have 
the capability to achieve and maintain subcriticality and long-term 
cooling. The requirements would be included to address the potential 
that some reactor designs may be able to achieve a stable end state for 
the purpose of event analyses but might need further actions to 
completely shut down and service the facility.
     Sec.  53.440(i): The design, analysis, and development of 
programmatic controls under part 53 would consider the number of 
reactor units and other significant inventories of radioactive 
materials contributing to the risks to public health and safety. This 
would reflect the definition of ``Commercial nuclear plant'' in subpart 
A and reinforce that the evaluation of LBEs is performed on a plant-
wide basis. This aspect of part 53 would be different from parts 50 and 
52, which generally define safety requirements on the assumption of 
events involving only individual reactor units.
     Sec.  53.440(j): A design requirement is proposed to 
provide a technology-inclusive requirement that would be equivalent to 
the requirements in Sec.  50.150 to address the possible impact of a 
large commercial aircraft.
     Sec.  53.440(k): The inclusion of a specific proposed 
requirement to address the risks to public health from potential 
chemical hazards of licensed material is appropriate given the 
diversity of reactor technologies and designs that might be licensed 
under part 53. The requirement in part 53 would be similar to the 
existing requirements in 10 CFR part 70, ``Domestic Licensing of 
Special Nuclear Material,'' that address both potential radiological 
and chemical hazards for licensed materials at fuel cycle facilities.
     Sec.  53.440(l): Provisions are proposed to require that 
measures be taken during the design of commercial nuclear plants to 
minimize contamination of the facility and the environment, facilitate 
eventual decommissioning, and minimize the generation of radioactive 
waste in accordance with Sec.  20.1406.
     Sec.  53.440(m): A design requirement is proposed to 
provide a technology-inclusive equivalent to the requirements in Sec.  
50.68 by including options for commercial nuclear plants to either have 
a monitoring system capable of detecting a criticality as described in 
Sec.  70.24 or to have restrictions on SNM handling and storage that 
would prevent inadvertent criticality events.
     Sec.  53.440(n): The design would need to reflect state-
of-the-art human factors principles for safe and reliable performance 
in all settings that human activities are expected for performing or 
supporting the continued availability of plant safety or emergency 
response functions.
    Section 53.450 would establish analysis requirements and would 
center upon the use of a PRA in combination with other generally 
accepted approaches for systematically evaluating engineered systems. 
The reliance on PRAs as a key component in the proposed analysis 
requirements for part 53 would reflect the decades of improvements in 
PRA methodologies and the increasing use of PRA techniques in the 
design, licensing, and oversight of both operating and future nuclear 
reactors. Part of the Commission's PRA Policy Statement is that the use 
of PRA technology should be increased in all regulatory matters to the 
extent supported by the state of the art in PRA methods and data and in 
a manner that complements the NRC's deterministic approach and supports 
the NRC's traditional defense-in-depth philosophy. The need to 
supplement PRA insights with other engineering approaches and judgments 
reflects the NRC's longstanding policy described in the SRM to SECY-98-
144, ``Staff Requirements--SECY-98-144--White Paper on Risk-Informed 
and Performance-Based Regulations,'' dated February 24, 1999, for 
regulatory decision-making to be risk-informed but not solely based on 
numerical results of a risk assessment (i.e., not a risk-based 
approach). Part 53 would maintain a role for NRC's traditional 
deterministic approaches (particularly for DBAs) and defense-in-depth 
philosophy by including specific requirements utilizing these 
regulatory tools in subparts B and C.
    PRA would be used in combination with other techniques in part 53 
to identify and categorize LBEs, classify SSCs, and evaluate defense in 
depth. This increased role for the PRA necessitates that it would be 
developed, performed, and maintained in accordance with NRC-approved 
standards and practices (see Sec.  53.450(c) and (d)). The computer 
codes used to model the plant response and the behavior of the barriers 
to the release of radionuclides would need to be qualified for the 
range of conditions being simulated across a wide range of unplanned 
events. These analyses would need to use realistic approaches and 
address uncertainties associated with states of knowledge, modeling, 
and performance of SSCs.
    While industry consensus PRA standards and peer review processes 
endorsed in RGs 1.200 and 1.247 remain

[[Page 86929]]

acceptable for developing a PRA, they are not regulatory requirements 
and an application under part 53 need not follow every aspect of the 
applicable consensus PRA standard. Existing processes for defining the 
scope and capability of a PRA supporting an application offer 
flexibility in determining the degree to which the PRA needs to be 
developed and may be informed by other factors such as design 
complexity and the needed degree of realism and level of detail, 
consistent with the use of the PRA and substance of the application. 
Such processes are currently available for appropriately defining the 
scope of the PRA and determining applicability of supporting 
requirements in consensus PRA standards needed to satisfy the proposed 
regulatory requirements for the specific uses of analyses under Sec.  
53.450(b). Likewise, NRC determinations of the acceptability of such 
PRAs would include consideration of the appropriateness of the 
applicant-defined scope as part of determining the applicability of and 
conformance to consensus PRA standard supporting requirements 
consistent with the current state of practice. In addition, these 
determinations would include consideration of other aspects of the 
development of the PRA, such as PRA peer reviews. An NRC determination 
of the acceptability of a PRA includes but is not limited to assessing 
the initial and boundary conditions and key assumptions used in the 
analysis, treatment of uncertainties, and the use of screening tools 
and bounding or simplified methods for any mode or hazard, provided the 
use of those tools and methods is justified by an acceptable technical 
basis. In that regard, the consensus PRA standards would not be applied 
by the NRC as a strict checklist of requirements for part 53 PRA 
acceptability determinations.
    The proposed Sec.  53.450(c) would require periodic maintenance and 
upgrading of the PRA to maintain an alignment between the supporting 
analyses and the design and performance of plant equipment, programs 
and procedures, and other factors associated with meeting the safety 
criteria of the proposed Sec.  53.220 and the evaluation criteria of 
proposed Sec.  53.450(e)(2). The periodic maintenance of the PRA would 
also be a means to consider new or revised information related to 
external hazards, industry operating experience, performance issues 
with or degradation of SSCs, and other contributors to the frequency 
and potential consequences of various event sequences. The periodic 
assessments performed by licensees to support the maintenance of the 
PRA and other requirements in the proposed part 53 would be 
complemented by NRC inspections and programs to assess new or revised 
information related to topics such as natural hazards, operating 
experience, and potential generic safety issues.
    The categories of LBEs used in part 53 would include anticipated 
event sequences, unlikely event sequences, and very unlikely event 
sequences. The unlikely event sequences would include those events with 
estimated frequencies well below the frequency of events expected to 
occur during the lifetime of a commercial nuclear plant. An important 
aspect of the analysis requirements is that, under proposed Sec.  
53.450(e), the analyses of LBEs other than DBAs would not only be used 
to show the performance criteria of Sec.  53.220 are satisfied but to 
also show that evaluation criteria defined for each LBE or category of 
LBEs would also be satisfied. Such evaluation criteria for specific 
LBEs or categories of LBEs would be defined in terms of limits on the 
release of radionuclides or maintaining the integrity of one or more 
barriers used to limit the release of radionuclides and reflect a 
graded approach of allowing lesser potential consequences from more 
frequent events. An example of such evaluation criteria for a range of 
LBEs that could likely be expanded for part 53 is provided in RG 1.233. 
Another proposed requirement for the proposed Sec.  53.450(e) analyses 
is that the methodology would need to include a means to identify event 
sequences deemed risk-significant such that those event sequences can 
be given special attention within other sections of part 53.
    Part 53 would maintain an important role for a deterministic 
analysis of DBAs in the performance criteria of Sec.  53.210 and the 
related analytical requirements in Sec.  53.450(f). The analysis of 
DBAs would be required to address event sequences drawn from those with 
estimated frequencies below the expected lifetime of a generation of 
reactors (e.g., event sequences with frequencies as low as one in ten 
thousand years). As proposed in this section, DBAs would need to be 
analyzed using deterministic methods and ensure a safe, stable end 
state with reliance upon only SR SSCs and human actions, if needed, to 
be performed by operators licensed under the provisions of Sec. Sec.  
53.760 through 53.795.
    While the DBAs analyzed under part 53 would be similar to the 
traditional DBAs analyzed under parts 50 and 52, there are important 
distinctions between the overall role of DBA analyses in part 50 and 
proposed part 53. In part 53, the role of the DBA analysis would be 
more narrowly focused on selecting SR SSCs and determining functional 
design criteria for those SSCs to ensure the commercial nuclear plant 
meets the safety criteria in Sec.  53.210. The overall control of risks 
posed by commercial nuclear plants under part 53 would be provided by 
the analyses of and measures taken for both DBAs and other LBEs, 
including very unlikely event sequences. This would contrast with the 
traditional deterministic approach in part 50 wherein the analyses of 
DBEs such as DBAs were used to provide bounding assessments, 
incorporate standard design rules such as assumptions related to single 
failures, and to define conservative performance requirements for SR 
SSCs. Limitations related to the traditional deterministic approach 
were addressed in part 50 through case-by-case assessments and specific 
actions for beyond-design-basis events such as anticipated transients 
without scram and station blackout.
    Section 53.450 would also include provisions to ensure that 
analyses are performed to support the design requirements of Sec.  
53.440(e) on fire protection, Sec.  53.440(j) on aircraft impact 
assessments, and Sec.  53.425 on using design features and plant 
programs to control doses to members of the public from routine 
effluents and direct radiation from contained sources. The proposed 
analysis requirements related to fire protection would support either a 
traditional, deterministic approach or a more risk-informed approach 
where the risks from fires are addressed within the identification and 
analyses of LBEs.
    Section 53.460 would establish criteria for the safety 
classification of SSCs and determination of appropriate special 
treatments. As noted in subpart A, the term ``Special treatments'' 
would be defined to mean those items, such as measures taken to satisfy 
functional design criteria, quality assurance, and programmatic 
controls, which provide assurance that certain SSCs will provide 
defense in depth or perform risk-significant functions. These 
requirements would also provide confidence that the SSCs will perform 
under the service conditions and with the reliability credited in the 
analysis performed in accordance with Sec.  53.450 to satisfy the 
safety criteria in Sec. Sec.  53.210 and 53.220. The terminology used 
in part 53 would include the following categories for SSC 
classification: (1) SR; (2) NSRSS; and (3) non-safety significant. 
Requirements for SR SSCs would be defined in other sections of

[[Page 86930]]

part 53 and would include using TSs for controls during operation and 
the application of quality assurance requirements from appendix B of 
part 50.
    Requirements for NSRSS SSCs would include the need to identify 
necessary special treatments such as performance measures on 
reliability. Licensees would generally be afforded flexibility in 
maintaining and changing special treatments for SSCs categorized as 
NSRSS. Non-safety-significant SSCs would be addressed under normal 
licensee programs for commercial grade equipment and typical industry 
practices for general plant design and maintenance. Safety-related SSCs 
would also contribute to defense in depth and risk-significant 
functions and may warrant special treatments beyond those defined for 
their SR functions to reflect their role in meeting the safety criteria 
in Sec.  53.220 and the evaluation criteria in Sec.  53.450(e).
    Section 53.470 would allow an applicant or licensee to seek 
operational flexibilities by adopting more restrictive criteria than 
those provided in Sec.  53.220 and that might otherwise be used in the 
analysis of LBEs under Sec.  53.450(e). Such an approach might be taken 
to ensure sufficient safety margins to gain operational flexibilities 
in areas such as justifying siting in relation to population centers or 
staffing levels. As an example, an applicant or licensee could propose 
to justify siting proposals by adopting alternate criteria for very 
unlikely event sequences. Such alternate criteria could require 
calculated consequences for an individual at the exclusion area 
boundary to be less than one rem total effective dose equivalent 
(TEDE). This section would establish requirements to ensure that, if 
more restrictive evaluation criteria than those required by a 
methodology were used to justify operational flexibilities, then the 
analysis, design features, and programmatic controls would be 
established and maintained accordingly.
    Section 53.480 would establish seismic design considerations. This 
proposed section would relate to the safety criteria in subpart B, the 
analytical requirements related to external hazards in Sec.  53.450, 
and subpart D, ``Siting Requirements.'' For licenses issued under part 
53, this section in subpart C would support a variety of approaches to 
seismic design. For example, a design for a commercial nuclear plant 
could show that SSCs are able to withstand the effects of earthquakes 
by adopting an approach similar to that in appendix S to part 50. 
Alternatively, an applicant could follow the more recent risk-informed 
alternatives afforded by standards development organizations (e.g., 
American Society of Civil Engineers (ASCE)/Structural Engineering 
Institute (SEI) 43-19, ``Seismic Design Criteria for Structures, 
Systems, and Components in Nuclear Facilities.'') Because the agency 
has not endorsed ASCE/SEI-43-19, an applicant can propose to use ASCE/
SEI 43-19 on an application specific basis to meet Sec.  53.480 and the 
NRC would evaluate the adequacy of the standard as applied in that 
application. The design could also be done with the full integration of 
seismic PRAs into the design and licensing of a particular commercial 
nuclear plant. This section has been developed to accommodate a variety 
of potential risk-informed, performance-based seismic design 
approaches. The analyses required by Sec.  53.450 would need to address 
seismic hazards as well as other external hazards. The expected 
responses of SSCs to a range of seismic events would be included in the 
analyses when ensuring that the safety criteria defined under Sec.  
53.220 would be met. The potential SSC responses to seismic hazards 
could be addressed in the analyses using a fragility model (conditional 
probability of its failure at a given hazard input level), a high 
confidence of low probability of failure value, or other method 
endorsed or otherwise found acceptable by the NRC.

Subpart D--Siting Requirements

    Proposed subpart D in part 53 would state requirements for the 
siting of commercial nuclear plants and would serve the role provided 
by 10 CFR part 100, ``Reactor Site Criteria,'' for nuclear reactors 
licensed under parts 50 and 52. As reflected in proposed Sec.  53.500, 
the reason for establishing siting requirements would remain the same 
as it has been historically, which is to ensure that licensees and 
applicants assess what impact the site environs may have on a 
commercial nuclear plant (e.g., external hazards) and, conversely, what 
potential adverse health and safety impacts a commercial nuclear plant 
may have on nearby populations in view of the site characteristics.
    Proposed Sec.  53.510 would require that design-basis external 
hazard levels be identified and characterized based on site-specific 
assessments of natural and constructed hazards with the potential to 
adversely affect plant functions. The site-specific assessments would 
be used in the proposed Sec.  53.415, which would require that SR SSCs 
be designed to withstand the effects of natural phenomena and 
constructed hazards of levels or severities up to design-basis external 
hazard levels. The design-basis levels for external hazards relevant to 
a site would need to account for uncertainties and variabilities in 
data, models, and methods used to characterize those hazards. Existing 
approaches could be used to demonstrate compliance with this 
requirement. The historical importance of assessing seismic events as 
risks to commercial nuclear plants and the associated development of 
risk-informed approaches to address seismic events would be reflected 
in proposed Sec.  53.480, ``Earthquake engineering,'' and specific 
requirements in subpart C. The NRC is developing a graded approach for 
seismic design by grouping SSCs into different seismic design 
categories (SDCs) based on their risk significance. While the agency 
has not endorsed ASCE/SEI-43-19, an applicant can propose to use ASCE/
SEI 43-19 on an application-specific basis to meet Sec.  53.480 and the 
NRC will evaluate the adequacy of the standard as applied in that 
application. The NRC staff will continue to review ASCE/SEI-43-19 as 
part of its efforts to further develop guidance in this area. The 
approach described in RG 1.208, ``A Performance-Based Approach to 
Define the Site-Specific Earthquake Ground Motion,'' would be an 
acceptable way to develop site-specific ground motion response spectra 
for SSCs under appendix S to part 50, which corresponds to SSCs that 
are categorized as the highest SDC (SDC-5) in ASCE/SEI 43-19.
    The evaluation of seismic hazards under subpart D would need to be 
sufficient to inform a site-specific design (e.g., a CP or custom COL) 
or confirm the use of a standard design for a commercial nuclear plant 
under Sec.  53.480 and other sections of subpart C. A risk-informed 
approach could use several design-basis ground motions (DBGMs) to 
assess SSCs in various SDCs (i.e., one DBGM per SDC). Section 53.510(d) 
would state that geologic and seismic siting factors must also include 
related hazards such as seismically induced flooding and volcanic 
activity that may affect the design and operation of a proposed 
commercial nuclear plant for the proposed site.
    Section 53.520 would require applicants to identify and assess site 
characteristics related to topics which might include meteorology, 
geology, hydrology, or other areas in the design and analyses required 
under subpart C.
    Proposed section 53.530 would set requirements for population-
related considerations and maintain requirements and definitions 
similar to those currently in part 100 for an exclusion area, low 
population zone,

[[Page 86931]]

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

Subpart E--Construction and Manufacturing Requirements

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

    \1\ On December 17, 1982, the NRC issued ``Manufacturing License 
ML-1 to Offshore Power Systems for the manufacture of a maximum of 
eight floating nuclear plants,'' dated September 30, 1982, but the 
project was subsequently canceled.
---------------------------------------------------------------------------

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

[[Page 86932]]

reactors onsite by a variety of contractors and subcontractors.
    Subpart E would include requirements that would apply to portions 
of a manufactured reactor in recognition that some activities covered 
by an ML may occur at different fabrication facilities. As with the 
preceding sections on construction, Sec.  53.620 would establish the 
requirements to have in place programs, procedures, and a well-defined 
command and control structure to manage manufacturing-related 
activities.
    Section 53.620(b) in subpart E would propose requirements for 
executing the manufacturing activities following receipt of an ML under 
part 53. Information about the design and manufacturing processes 
should be provided by the applicant. The importance of the ML is 
reflected in several of the proposed requirements in Sec.  53.620(b) 
that would refer to complying with the ML, including conducting 
manufacturing processes within facilities for which the license holder 
can control activities. The essential role of post-manufacturing 
inspections would also be incorporated into this proposed section by 
requiring the holder of the ML to perform inspections and have 
acceptance processes for manufactured reactors or portions of a 
manufactured reactor.
    Section 53.620(c) would provide proposed requirements for the 
control of radioactive materials if the holder of an ML plans to 
possess and use source, byproduct, or SNM as part of the manufacturing 
process. By and large, the proposed subpart E would refer to NRC 
regulations in 10 CFR part 30, ``Rules of General Applicability to 
Domestic Licensing of Byproduct Material,'' 10 CFR part 40, ``Domestic 
Licensing of Source Material,'' and part 70 for the requirements on 
controlling radioactive materials. Several specific requirements to 
address the potential hazards of radioactive materials are proposed in 
areas such as having a fire protection program, an emergency plan, 
training programs, and procedures to minimize contamination.
    The most significant change proposed for MLs in part 53 as compared 
to MLs under part 52 relates to Sec.  53.620(d) in subpart E and the 
associated licensing provisions in subpart H. These provisions would 
allow and establish requirements for the loading of fuel into a 
manufactured reactor at the manufacturing site for subsequent transport 
to a commercial nuclear facility that will operate pursuant to a COL. 
The first requirement in the proposed Sec.  53.620(d) would establish 
limitations on when a license under part 70 would authorize the loading 
of fuel into a reactor manufactured under an ML. The proposed 
regulation would require the manufactured reactor to include at least 
two independent physical mechanisms that will each prevent criticality 
should conditions most favorable to critical operation be introduced 
(e.g., optimum neutron moderation and reflection). This requirement 
would contribute to the NRC's longstanding practice of requiring 
defense in depth for preventing accidents in any facility dealing with 
SNM, including requirements in Sec.  70.64 for certain part 70 
licensees to adhere to the ``double contingency principle.''
    The requirements to have in place mechanisms to prevent criticality 
could likewise support meeting other provisions in subpart H to part 
70, such as those related to having a safety program and integrated 
safety assessment. The mechanisms to preclude criticality in the 
proposed requirements would reasonably ensure that a manufactured 
reactor would not become critical assuming optimum neutron moderation, 
and optimum neutron reflection conditions. With the proposed 
requirements for mechanisms to prevent criticality and all criticality 
safety controls required by 10 CFR part 70 in place, the presence of 
fuel in the manufactured reactor would not create a nuclear hazard 
different than the hazard from the presence of the same fuel in a 
storage location or container licensed under 10 CFR part 70. 
Collectively, the proposed measures would reasonably ensure that the 
manufactured reactor would not be capable of operations, thereby 
obviating the need for a COL under Sec. Sec.  53.1416 and 53.1440 to 
authorize fuel loading. Additionally, this approach would focus the ML 
application and its review on the design, manufacture, and deployment 
of the manufactured reactor.
    The activities involving SNM within the manufacturing facility, 
including the loading of fuel, would be regulated primarily under the 
part 70 license. The reference to the requirements in subpart H of part 
70 in section 53.620(d) assures that the activities involving the 
receipt, storage, and loading of a variety of possible fuel forms and 
enrichments at the manufacturing facility will be analyzed in a 
systematic manner and appropriate protection will be provided against 
equipment malfunctions, human errors, external hazards, and other 
adverse conditions. The regulations in part 51 provide a flexible 
approach for environmental review to address the range of regulated 
activities under part 70. The flexibility in part 51 will enable the 
NRC to determine the appropriate type of environmental review based on 
the circumstances associated with the loading of fuel into a specific 
manufactured reactor.
    The proposed Sec.  53.620(d) cites the requirements in parts 70, 
71, and 73 to ensure important features and programs are in place prior 
to the receipt of SNM. The features and programs required to be in 
place prior to receipt of SNM include (1) radiation monitoring 
instrumentation and alarms; (2) measures to detect potential 
criticality accidents; (3) appropriate procedures, equipment, and 
personnel qualified for the fuel loading; (4) programs for physical 
security and cybersecurity; and (5) material control and accounting 
(MC&A) programs. Section 53.620(d)(2)(i) proposes requirements to 
address security programs for any ML authorizing possession of a 
manufactured reactor into which fuel has been loaded at the 
manufacturing facility. Currently, for category II SNM, security 
measures may be required in addition to requirements included in Sec.  
73.67, ``Licensee fixed site and in-transit requirements for the 
physical protection of special nuclear material of moderate and low 
strategic significance,'' on a case-by-case basis. Including 
appropriate security measures in the proposed part 53 regulations will 
provide additional openness and transparency for applicants applying 
for an ML who seek to load fuel into manufactured reactors at a 
manufacturing site.
    Currently, Sec.  73.67 only requires a security plan for licensees 
who possess, use, transport, or deliver to a carrier for transport SNM 
of moderate strategic significance, or 10 kg or more of SNM of low 
strategic significance. However, the proposed physical security program 
for fueled manufactured reactors would require a security plan for any 
ML authorizing possession of a manufactured reactor into which fuel has 
been loaded at the manufacturing facility, regardless of fuel type, 
enrichment, and quantity. This is consistent with other controls for 
MLs, including reactivity and criticality controls.
    The proposed requirements would also require a holder of an ML and 
part 70 license to address cybersecurity to ensure a cyberattack would 
not adversely impact the functions performed by digital assets used by 
the licensee for physical security, radiation monitoring, or 
criticality prevention.
    The proposed regulations in part 53 covering the activities related 
to the storage, movement, and loading of fresh

[[Page 86933]]

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

Subpart F--Requirements for Operation

    Proposed subpart F would provide the requirements for the 
operations phase of a commercial nuclear plant to ensure that the 
safety criteria in subpart B are satisfied throughout the plant's 
lifetime and during all modes of normal operation and unplanned events. 
Section 53.700 would provide the overall objectives and general 
organization of subpart F, which would be to establish requirements 
during operations for: (1) plant SSCs; (2) plant personnel; and (3) 
plant programs.
    Proposed Sec.  53.710 would provide the requirements for 
maintaining capabilities, availability, and reliability of SSCs to 
demonstrate compliance with the safety criteria and design requirements 
for unplanned events that are described in proposed subparts B and C. 
The basic structure of this proposed section would be that controls for 
SR SSCs are provided by TS and controls for NSRSS SSCs are required to 
be addressed with licensee-controlled documents and procedures.
    The general content and control of TS under the proposed part 53 
would be similar to the requirements in part 50. The proposed 
requirements for TS would include limits on the inventories of 
radioactive materials, plant operating limits, and specific 
requirements for each SR SSC, including limiting conditions for 
operation (LCO) and required surveillances. The proposed requirements 
for TS would also include a section on important design elements, which 
is similar to design features in Sec.  50.36, and a section for 
administrative controls. A provision addressing the development and 
submittal of TS to address decommissioning activities would also be 
included in the proposed subpart G.
    The proposed requirements for TS under part 53 would not carry over 
safety limits or associated limiting safety system settings from Sec.  
50.36, which contains TS requirements for operating reactors under 
parts 50 and 52. As discussed in SECY-18-0096, systematic assessments 
and more mechanistic approaches to evaluating source terms support an 
alternative approach to establishing barrier-based safety limits. An 
example provided in that paper is a comparison of: (1) the traditional 
specified acceptable fuel design limits (SAFDL) that support protecting 
a specific barrier from potential failure mechanisms (e.g., departure 
from nucleate boiling to protect fuel cladding); and (2) the specified 
acceptable system radionuclide release design limit (SARRDL) concept, 
which limits the possible increase in circulating radionuclide 
inventory during normal operations or an AOO as part of an integrated 
or ``functional containment'' approach. Additional discussion of the 
use of SARRDL in the design and licensing of advanced reactors is 
provided in RG 1.232. The SARRDL could be addressed as an operating 
limit within this proposed construct of requirements for TS. In cases, 
such as LWRs, where a SAFDL approach might be used as part of a 
mechanistic approach to meeting the design and analysis requirements in 
subpart C, the associated functional design criteria proposed in Sec.  
53.410 and TS under the proposed Sec.  53.710(a) would define similar 
requirements as those provided by the safety limit and limiting safety 
system setting requirements in Sec.  50.36.
    The proposed requirements for TS under part 53 would not include 
specific criteria for identifying when LCOs must be established (i.e., 
would not include an equivalent to Sec.  50.36(c)(2)(ii)). Instead, 
consistent with subparts B and C, the TS requirements in subpart F of 
part 53 would define TS LCOs as providing limits on SR SSCs. The SR 
SSCs protect against DBAs to demonstrate compliance with the safety 
criteria in the proposed Sec.  53.210. In the proposed construct for 
part 53, risk-significant SSCs would be addressed through a combination 
of TS for the SR SSCs and establishment and monitoring of performance 
standards for NSRSS SSCs.
    In addition to addressing TS for SR SSCs, proposed Sec.  53.710 
would require appropriate controls be developed and implemented for 
NSRSS SSCs. Examples include appropriate surveillances and controls 
established through reliability assurance programs. Configuration 
management and other special treatments would provide that the 
capabilities, availabilities, and reliabilities of NSRSS SSCs are 
maintained consistent with the underlying risk assessments while 
providing flexibility to licensees through maintaining the management 
functions within licensee-controlled programs. Controls on NSRSS SSCs 
are appropriate as part of the overall performance-based approach 
within proposed part 53. Special treatments beyond those defined for 
their SR functions may also be warranted for SR SSCs to reflect their 
role in meeting the safety criteria in Sec.  53.220 and the evaluation 
criteria in Sec.  53.450(e). The performance objectives for NSRSS SSCs 
would reflect that the comprehensive risk metrics and related risk 
performance objectives established under Sec.  53.220 may involve 
assessing

[[Page 86934]]

and averaging the risks over a defined period (e.g., plant year) and 
would not constitute a real-time requirement that must be continuously 
demonstrated by the licensee. The controls under Sec.  53.710(b) 
justify proposed changes in part 53 from the traditional or 
deterministic approaches in parts 50 and 52 in areas such as replacing 
the single-failure criterion with a probabilistic reliability criterion 
(see SRM-SECY-03-0047, ``Policy Issues Related to Licensing Non-Light-
Water Reactor Designs,'' dated June 26, 2003). This approach could also 
support the incorporation of risk insights and analytical margins to 
gain operational flexibilities in areas such as siting and staffing 
requirements described in subsequent sections of proposed subpart F.
    Proposed Sec.  53.715 would provide the requirements for developing 
and implementing a program to do the following: (1) control maintenance 
activities; (2) take appropriate corrective action when performance 
issues are identified; (3) conduct routine evaluations of 
effectiveness; and (4) assess and manage risks resulting from 
maintenance activities. These proposed requirements are similar to 
those included in Sec.  50.65 (maintenance rule), including the need to 
assess and manage the increase in risk that may result from the 
proposed maintenance activities. While, for the maintenance rule, 
specific criteria must be developed to capture both SR and non-SR but 
otherwise important SSCs, the proposed Sec.  53.715 would cover SR SSCs 
and NSRSS consistent with other subparts in part 53.
    Proposed Sec.  53.720 would provide the requirements for responding 
to a seismic event during the operating phase of the life cycle of a 
commercial nuclear plant and would be equivalent to the requirements in 
paragraph IV(a)(3) of appendix S, ``Earthquake Engineering Criteria for 
Nuclear Power Plants,'' to part 50.
    The proposed part 53 would include provisions to address staffing, 
training, personnel qualifications, and human factors engineering (HFE) 
in a manner that is risk informed, technology inclusive, performance 
based, and flexible in nature. During the development of part 53, the 
staff prepared a draft white paper on ``Risk Informed and Performance 
Based Human-System Considerations for Advanced Reactors,'' to support 
interactions with stakeholders and the ACRS. Key considerations include 
the recognition that staffing, operator qualifications, and HFE are 
interconnected areas that must be approached in an integrated manner 
and, furthermore, that safety functions, including the means by which 
they are fulfilled, provide an effective method for informing 
technology-inclusive requirements.
    The requirements associated with this approach would be in 
Sec. Sec.  53.725 through 53.830. Section 53.725 discusses 
applicability and defines specific terms. Some definitions draw from 
those in Sec.  55.4. Several new definitions would be introduced for 
use within the context of subpart F. These new definitions would be the 
following: ``Automation,'' ``Auxiliary operator,'' ``Generally licensed 
reactor operator,'' ``Interaction-dependent-mitigation facility,'' 
``Load following,'' ``Self-reliant-mitigation facility.''
    Sections 53.725 through 53.830 would be divided into four portions 
that would cover general operational requirements, operator and senior 
operator licensing requirements, generally licensed reactor operator 
(GLRO) requirements, and general training requirements for plant staff. 
The NRC intends to provide guidance addressing the review of operator 
staffing plans; the review of operator, senior operator, and GLRO 
examination programs; and the implementation of scalable HFE reviews. 
Licensees would be required to use GLROs upon demonstrating compliance 
with the criteria in Sec.  53.800.
    Certain routine communications are necessary to facilitate the 
operator licensing process. The NRC is proposing to adapt the 
requirements of Sec. Sec.  55.5 and 50.74 to Sec.  53.726 to accomplish 
this.
    Specific information must be collected in order to facilitate the 
initial issuance of operator licenses, as well as to allow for license 
renewals and required updates thereafter. Such information collection 
activities must also be approved by the OMB. The NRC is proposing to 
adapt the requirements of Sec.  55.8, to include any needed updates in 
OMB approval information, to Sec.  53.120 to accomplish this.
    The information used within the regulatory processes of the NRC 
must be free from omissions and inaccuracies to facilitate effective 
regulation. Consistent with this, the NRC is proposing to adapt the 
requirements of Sec.  55.9 to Sec.  53.728 to require the completeness 
and accuracy of material information provided by individual applicants 
and license holders.
    Section 53.730 would provide performance-based and technology-
inclusive requirements for assessing the role of personnel in facility 
safety, applying human-system considerations within facility design, 
and incorporating operational approaches that are consistent with 
design-specific safety considerations. Most of these requirements would 
be adapted from portions of Sec. Sec.  50.34(f) and 50.54 and 10 CFR 
part 55, ``Operators' Licenses,'' with considerable modification in 
order to reflect the introduction of new technologies and possible 
changes in the roles of personnel in preventing and mitigating events. 
The NRC is proposing that these technical requirements would, together, 
serve as a component of the required content of applications for OLs 
and COLs under part 53. Additionally, the NRC proposes that the 
specific technical requirements associated with HFE, human-system 
interface design, concept of operations, functional requirements 
analysis, and function allocation would serve as a component of the 
required content of applications for standard DCs, standard design 
approvals, MLs, and CPs, as well.
    Human factors engineering is essential to facilitate the role of 
personnel in facility safety in a manner that is both effective and 
reliable. The NRC proposes to adapt Sec.  53.730(a) from the HFE design 
requirements of Sec.  50.34(f)(2)(iii). A key difference would be that 
the requirement would now be focused on settings where personnel 
fulfill their safety or emergency response roles wherever they may 
occur. The NRC additionally proposes to include within the scope of 
this requirement activities for assuring the continued availability of 
plant equipment that is needed for safety, and envisions that this may 
encompass relevant maintenance, inspections, and testing as well. The 
NRC intends that this requirement would be associated with staff 
guidance for conducting scalable reviews of HFE that is planned to 
accompany part 53.
    Human-system interfaces provide vital information to operators 
across a spectrum of operating conditions that can range from normal 
operations through severe accident conditions. The specific types of 
information that must be available to support operations staff during 
such conditions include, in part, those associated with safety function 
parameters, safety system status, possible core damage states, barrier 
integrity, and radioactive leakage. Due to the importance of such 
information, the NRC proposes under Sec.  53.730(b) to require such 
human-system interface design features for all facilities, irrespective 
of other flexibilities proposed under part 53. Therefore, the NRC 
proposes to adapt specific post-Three Mile Island requirements of Sec.  
50.34(f) in a technology-inclusive manner as detailed in the following:

[[Page 86935]]

     Paragraph (b)(1) would be adapted from Sec.  
50.34(f)(2)(iv).
     Paragraph (b)(2) would be adapted from Sec.  
50.34(f)(2)(v).
     Paragraph (b)(3) would be adapted from Sec.  
50.34(f)(2)(xi), 50.34(f)(2)(xii), and 50.34(f)(2)(xxi).
     Paragraph (b)(4) would be adapted from Sec.  
50.34(f)(2)(xvii), 50.34(f)(2)(xviii), 50.34(f)(2)(xix), and 
50.34(f)(2)(xxiv).
     Paragraph (b)(5) would be adapted from Sec.  
50.34(f)(2)(xxvi).
     Paragraph (b)(6) would be adapted from Sec.  
50.34(f)(2)(xxvii).
    In addition to the requirements of Sec.  53.730(b)(1) through (6), 
a further set of human-system interface design requirements applicable 
only to those facilities that will be staffed by GLROs would be 
provided under Sec.  53.730(b)(7). This prescriptive set of design 
requirements for those facilities which demonstrate compliance with the 
criteria of Sec.  53.800 would recognize that the application of HFE 
under Sec.  53.730(a) is anticipated to be significantly reduced at 
such facilities in the absence of an expected operator role for the 
fulfillment of safety functions. However, it should be noted that the 
capability for an immediately initiated, manual reactor shutdown would 
be conservatively mandated irrespective of any other design 
considerations.
    The NRC proposes Sec.  53.730(c) to require the submittal of a 
concept of operations that is of sufficient scope and detail to 
appropriately inform the staff. The development of a concept of 
operations can facilitate a clear understanding on the part of the NRC 
for potential novel operating concepts. Additionally, such information 
is likely to reduce the degree of resources and interactions needed for 
the NRC to obtain the understanding necessary to enable flexible 
requirements in areas such as staffing, operator qualifications, and 
HFE.
    The NRC proposes Sec.  53.730(d) to require the submittal of both a 
Functional Requirements Analysis and a Function Allocation. The 
identification of design-specific safety functions and how they are 
fulfilled serves as a primary means for achieving technology-inclusive 
requirements within areas such as staffing, operator qualifications, 
and HFE. The Functional Requirements Analysis and Function Allocation 
processes (which are both HFE methods derived from systems engineering 
principles), provide an effective means to identify both how safety 
functions will be satisfied and how to characterize any associated 
operator role in doing so. A Functional Requirements Analysis shows 
what features, systems, and human actions are relied upon to 
demonstrate safety (i.e., fulfill safety functions). A Function 
Allocation then describes how safety functions are assigned to both 
personnel and automatic systems. However, an important adaptation of 
the Function Allocation for use under the proposed rule would be the 
further need to not only describe allocations of safety functions to 
human action and automation, but also to identify allocations made to 
active safety features, passive safety features, or inherent safety 
characteristics as well.
    Operating experience provides an important source of information by 
which to inform various aspects of facility design and operations. 
Accordingly, the NRC proposes in Sec.  53.730(e) to adapt the 
requirements of Sec.  50.34(f)(3)(i) for requiring an operating 
experience program.
    New technologies may involve concepts of operations that are more 
conducive to customizable licensed operator staffing requirements than 
the prescriptive requirements of Sec.  50.54(m). Analyses and 
assessments that are based on HFE principles provide a performance-
based means of determining licensed operator and senior operator 
staffing needed to support safe operations. In contrast, for those 
facilities required to be staffed by GLROs, the NRC anticipates that 
the operator staffing plans will reflect a simpler approach of showing 
that a continuity of responsibility will be maintained for facility 
operations throughout the operating phase, with at least one GLRO 
providing continuous oversight and remaining immediately available when 
any units are fueled. Additionally, a revised approach to the 
traditional position of the shift technical advisor that focuses on the 
availability of engineering expertise as a means of addressing 
uncertainties and abnormal circumstances is more suitable within the 
context of part 53 and is intended to be applicable to all facilities, 
irrespective of other design and staffing considerations.
    Consistent with this approach, the NRC proposes under Sec.  
53.730(f) to require the submittal of a staffing plan that details 
operations staffing, how engineering expertise will be provided, and 
what staffing will be available to provide other needed support 
functions. The NRC intends that this requirement would be associated 
with staff guidance for reviewing operations staffing plans that is 
planned to accompany part 53 and that, following NRC approval of the OL 
or COL, the staffing plan would become a condition of the facility 
license. The NRC intends that, at a minimum, the approved licensed 
operator and senior operator (or, if applicable, GLRO) staffing, 
positions, and personnel locations will be incorporated into 
corresponding requirements within the facility TS and that a license 
amendment would thus be required for any subsequent changes.
    Operator training and qualification programs provide an essential 
component of supporting human performance in implementing tasks with 
safety implications. Such programs must include components that cover 
the stages of initial training, examination, and continuing training. 
Additionally, recognizing the potential for varying concepts of 
operations to affect traditional, prescriptive approaches to operator 
proficiency, the NRC proposes under part 53 to allow facilities to 
develop operator proficiency programs based on facility-specific 
considerations.
    Therefore, the NRC proposes in Sec.  53.730(g)(1) to require 
approval as part of its approval of the OL or COL, of the programs that 
will be used for the initial training, initial examination, 
requalification training and examination, and proficiency of both 
licensed operators and senior operators. In a corresponding manner, the 
NRC proposes in Sec.  53.730(g)(2) to require approval of the programs 
that will be used for the GLRO equivalents of each of these programs 
for facilities with such staffing. The NRC intends that examination 
program requirements would be associated with staff guidance for the 
review of tailored examination processes that are planned to accompany 
part 53. Following the completion of an initial training program, 
continuing training programs provide an important means of sustaining 
the knowledge and abilities of individuals. The NRC is proposing to 
adapt the requirements of Sec.  50.54(i-1) in Sec.  53.730(g)(3) to 
require that operator continuing training programs be in effect to 
support operator performance. Under part 53, the NRC proposes to 
require these programs to be in effect concurrent with when the initial 
operator examinations first commence, in effect putting the programs in 
place only when they are needed. This represents a modification of the 
comparable requirement of Sec.  50.54(i-1), which links the 
commencement of these programs to a timeline driven by the licensing of 
the facility.
    The authorization to manipulate controls of the facility that 
directly affect reactivity or power level is restricted to individuals 
who are either licensed operators, licensed senior operators, or GLROs. 
However, for practical purposes, situations in which

[[Page 86936]]

an individual is participating in an approved training program or 
reestablishing proficiency may also call for them to operate the 
controls of the facility under the cognizance of a licensed individual. 
The NRC is proposing to adapt the requirements of Sec.  55.13 in Sec.  
53.735 to accomplish this, with a notable difference being the 
incorporation of GLROs.
    Section 53.740 would provide requirements for OL and COL holders 
under part 53. Portions of Sec.  53.740 would be adapted from the 
conditions of Sec.  50.54. In general, the conditions for operations 
staffing under part 53 would reflect considerations for potential 
technological differences and varying concepts of operation that are 
expected among part 53 facility licensees. Additionally, certain 
requirements would be specific to the operating phase while others 
would remain in effect following the permanent cessation of facility 
operations during the decommissioning phase.
    All commercial nuclear plants licensed under part 53 would require 
some form of licensed operator staffing, whether it be by specifically 
or generally licensed operators. Consistent with this, the NRC is 
proposing under Sec.  53.740(a) to require facility licensees to 
demonstrate compliance with the programmatic requirements for either 
specifically licensed operators and senior operators or for GLROs, as 
applicable to the facility.
    The NRC recognizes that technology-inclusive facility staffing will 
need to account for a potentially wide range of concepts of operations; 
for this reason, flexible and performance-based approaches for 
establishing required facility staffing are appropriate. However, once 
the appropriate facility staffing has been determined and approved by 
the NRC, such staffing must be maintained to ensure that the 
appropriately qualified individuals will be available when needed to 
support the safe operation of the facility. Therefore, the NRC is 
proposing under Sec.  53.740(b) to require that the staffing described 
within the approved facility staffing plan be maintained as a condition 
of the facility license as opposed to prescriptive staffing 
requirements like those of Sec.  50.54(k) and (m).
    Because operation of facility controls directly affects reactivity 
or power level, only those individuals who possess appropriate levels 
of qualification and authorization are permitted to operate those 
controls. The NRC is proposing to adapt the requirements of Sec.  
50.54(i) in Sec.  53.740(c) to require that only specifically licensed 
operators and senior operators or, alternatively, GLROs, may operate 
facility controls, with allowance for specified exceptions for the 
purposes of operator training or proficiency.
    Senior operators, by virtue of their license level, are qualified 
and authorized both to perform certain important responsibilities and 
to direct the licensed activities of licensed operators. Therefore, 
facilities that are required to be staffed by specifically licensed 
operators must also include senior operators within their staffing. In 
contrast, facilities staffed with GLROs only have a single license 
level available and, therefore, there is no equivalent provision for 
such facilities. The NRC is proposing to adapt the requirements of 
Sec.  50.54(l) in Sec.  53.740(d) to require the licensing and 
designation of senior operators at facilities staffed by specifically 
licensed operators.
    In contrast with control manipulations that directly affect reactor 
power and reactivity (e.g., control rod movement, control drum 
rotation, recirculation pump speed adjustment, reactor coolant system 
boration or dilution, etc.) and are therefore restricted to performance 
only by licensed operators, other types of plant operations that may 
result in reactor power and reactivity changes via means that are 
indirect in nature (e.g., electrical generation changes, turbine bypass 
valve operation, steam usage by process heat applications, etc.) may be 
implemented by non-licensed personnel. However, due to the potential 
influence of such operations on reactor power and reactivity, the 
continuous oversight of reactor parameters by a licensed operator is 
necessary during these operations. The NRC is therefore proposing to 
adapt the requirements of Sec.  50.54(j) in Sec.  53.740(e) to require 
appropriate oversight of operations, other than those associated with 
the controls themselves, that may affect reactivity or power level.
    Load following where plant output automatically changes in response 
to externally originated instructions or signals is not permitted under 
the existing regulations of Sec.  50.54. However, new technological 
considerations and concepts of operation may justify such an 
operational approach under appropriate circumstances. The NRC 
recognizes that, beyond electrical power generation, load following may 
also affect other applications of plant output, such as hydrogen 
production, desalination, or district heating. For load following to be 
permissible, measures must be in place to provide assurance that plant 
output considerations are not permitted to lead to challenges to safe 
reactor operations. These measures may consist of automated control 
systems, automatic protective features, or the continuous oversight and 
immediate intervention capability of an appropriately qualified and 
authorized individual. Section 53.740(f) would allow for load 
following, provided that appropriate measures are in place. In 
considering the acceptability of the measures associated with load 
following, the NRC expects that any automatic protection relied upon 
would be separate from that credited for reactor protection purposes 
and would employ setpoints that are set so as to prevent actuation of 
the reactor protection system while accomplishing its functions to the 
extent practical.
    Core alterations such as refueling are associated with specific 
considerations that warrant limiting the oversight of such operations 
to appropriately qualified and authorized individuals. Unlike other 
types of fuel handling operations, core alterations occur within the 
confines of a reactor vessel that is specifically designed to support 
and sustain nuclear criticality, thereby justifying the imposition of 
higher qualification levels within such contexts. The NRC is proposing 
to adapt the requirements of Sec.  50.54(m)(2)(iv) in Sec.  53.740(g) 
to require the supervision of core alterations by either a specifically 
licensed senior operator, a specifically licensed senior operator whose 
license is limited to fuel handling, or by a GLRO, as applicable to the 
facility. Because certain commercial reactor designs may be capable of 
refueling while at power and, in any event, overall facility oversight 
would already be required by either a specifically licensed senior 
operator or by a GLRO, the NRC proposes to omit this requirement as 
redundant during periods where core alterations occur while the plant 
is operating.
    It is impossible to predict every possible scenario that a 
commercial nuclear plant might potentially encounter. Therefore, it is 
prudent to grant the authority for appropriately qualified individuals 
to depart from facility license conditions when emergency circumstances 
dictate that doing so is in the interest of public health and safety. 
The NRC is proposing to adapt the requirements of Sec.  50.54(x) and 
(y) in Sec.  53.740(h) to permit specific individuals to authorize 
departures from facility license conditions or TSs when emergency 
conditions warrant doing so for the protection of the public health and 
safety. Recognizing that certain facilities licensed under part 53 may 
be staffed by GLROs in lieu of specifically licensed senior operators, 
the NRC proposes to extend this authority to

[[Page 86937]]

GLROs. While it is not anticipated that GLROs will have a role in the 
fulfillment of safety functions at self-reliant-mitigation facilities 
and, furthermore, that operators at such facilities would not be in a 
position by which to significantly influence radiological safety 
outcomes, the very nature of the Sec.  50.54(x) and (y) and the 
proposed Sec.  53.740(h) provisions concern situations that are 
unanticipated and, therefore, unforeseeable. Thus, it is appropriate to 
grant GLROs a comparable authority to that of senior licensed operators 
and certified fuel handlers as it relates to invoking this provision 
under emergency conditions as a means of accounting for such 
possibilities.
    Due to the unique authorities and responsibilities of both 
specifically and generally licensed reactor operators, it is essential 
that any individual fulfilling such a role demonstrate compliance with 
the regulatory requirements for operator licensing. Section 107 of the 
Act authorizes the Commission to prescribe conditions for the licensing 
of operators and to issue licenses consistent with those conditions. 
The NRC is proposing to adapt the requirements of Sec.  55.3 in Sec.  
53.745 to require that any person performing the function of an 
operator, senior operator, or GLRO must be authorized by a license 
issued by the Commission.
    The NRC proposes to license individuals as operators under both 
specific and general licensing frameworks. Specific licenses would be 
for licensed operators (i.e., reactor operators) and senior operators 
(i.e., senior reactor operators) and would be issued to a named person 
upon approval by the Commission of an application for that named 
person. In contrast, GLROs would perform duties under the provisions of 
a general license that would be effective without the filing of an 
application with the Commission or the issuance of licensing documents 
to a particular person. The NRC proposes requirements for the use of a 
specific licensing process for licensed operators and senior operators 
under Sec. Sec.  53.760 through 53.795, with Sec.  53.760 addressing 
applicability.
    Medical fitness is an important component of the overall process of 
specifically licensing operators because it provides assurance that 
operators will be able to carry out important duties without being 
precluded from doing so by health-related issues. Medical fitness also 
provides assurance that such issues will not adversely affect the 
performance of assigned job duties or cause operational errors that 
endanger public health and safety. In addition to a requirement for 
medical fitness, a medical examination by a physician to confirm 
compliance with this requirement is necessary. The NRC is proposing to 
adapt the requirements of Sec. Sec.  55.21, 55.23, and 55.27 under 
Sec.  53.765 to require medical fitness, examinations by physicians, 
and medical certification for specifically licensed operators and 
senior operators. In recognition of the fact that GLROs are not 
expected to have a role in the fulfillment of safety functions at the 
facilities at which they are licensed, the NRC proposes to not extend a 
comparable medical requirement to GLROs.
    The NRC is also proposing to adapt the requirements of Sec. Sec.  
55.25 and 50.74(c) in Sec.  53.770 to require that timely notifications 
be made to the NRC if a specifically licensed operator or senior 
operator develops a permanent physical or mental condition that 
adversely affects the performance of assigned operator job duties or 
could cause operational errors endangering public health and safety. 
Notwithstanding this requirement related to permanent medical 
conditions, the NRC continues to recognize that it is appropriate for 
facility licenses to impose administrative restrictions and conditions 
upon specifically licensed operators and senior operators in response 
to temporary medical conditions.
    The process of specifically licensing individuals as licensed 
operators or senior operators requires the submittal of applications to 
the NRC for review. These applications must detail certain elements 
associated with licensing, including the demonstration of compliance 
with examination, experience, and medical requirements. The NRC is 
proposing to adapt the requirements of Sec. Sec.  55.31 through 55.35 
in Sec.  53.775 to include requirements for the applications associated 
with the specific licensing of licensed operators and senior operators 
at commercial nuclear plants licensed under part 53. In contrast with 
the part 55 requirements, the NRC proposes to provide additional 
flexibility by locating certain details associated with the preparation 
and submittal of these applications within guidance in lieu of 
placement within this proposed rule itself.
    The NRC proposes overall programmatic requirements for specifically 
licensed operator and senior operator training, examination, and 
proficiency in Sec.  53.780. In general, the proposed requirements are 
adapted from those in part 55, with several additional flexibilities 
being incorporated to better account for potential variations in 
reactor technologies and concepts of operations. The requirements 
proposed in Sec.  53.780 cover, in part, the initial training, initial 
examination, requalification training, requalification examination, and 
proficiency of specifically licensed operators and senior operators.
    The initial training process provides individuals with the 
knowledge and abilities needed to subsequently fulfill assigned duties 
as licensed operators or senior operators in a safe and reliable 
manner. The use of a systems approach to training (SAT) ensures that 
the training program is based upon job requirements in a manner that 
can be adapted to account for differences in plant technology, concepts 
of operations, and operator roles in the fulfillment of design-specific 
safety functions. The NRC is proposing under Sec.  53.780(a) to require 
facility licensees to implement a SAT-based training program for the 
initial training of licensed operator and senior operator applicants. 
The program must be adequate to ensure that applicants will be capable 
of performing the duties necessary both to protect public health and 
safety and to maintain plant safety functions. The NRC further proposes 
that such programs be subject to NRC approval and subsequent change 
control processes of an appropriate nature.
    Examinations provide a means of assessing that individuals have 
achieved a degree of knowledge and ability that is sufficient to carry 
out assigned duties as licensed operators or senior operators in a 
manner that is safe and reliable. The NRC is proposing to adapt the 
requirements of Sec. Sec.  55.40, 55.41, 55.43, and 55.45 in Sec.  
53.780(b) to require that facilities establish and implement an initial 
examination program. However, a key difference from the comparable 
requirements of part 55 would be that facilities have the flexibility 
to propose, subject to NRC approval, the examination methods and 
criteria to be used in assessing satisfactory applicant performance. 
Such examination programs (including those used within the scope of 
requalification training) would need to provide for acceptable levels 
of both test validity and test reliability in order to be considered 
acceptable. The NRC intends that staff guidance would be available to 
facilitate the review of licensing examination programs that are 
proposed by facility licensees and that, following NRC approval, 
initial examination programs would be subject to an appropriate change 
control process. Furthermore, the NRC proposes that holders of licenses 
to operate commercial nuclear

[[Page 86938]]

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

 Reactor Theory, Thermodynamics, and Chemical Interactions
 Plant Systems and Components
 Reactivity Management and Manipulations
 Radiation Control and Safety
 Emergency, Abnormal, and Normal Operations
 Administrative Requirements and Conditions of the Facility 
License

    Requalification training programs provide for the continuing 
training and examination of specifically licensed operators and senior 
operators to ensure that they maintain the knowledge and abilities 
needed to support the safe and reliable performance of job duties 
following the completion of an initial training and examination 
program. The NRC is proposing to adapt the requirements of Sec.  55.59 
in Sec.  53.780(c) to require that facilities implement both a SAT-
based requalification training program and a biennial requalification 
examination program. However, a notable difference from the biennial 
requalification examinations required under part 55 would be that 
distinct annual operating test and biennial written examination 
components would not be mandated, with the facility licensee instead 
proposing the examination methods and criteria to be used in assessing 
satisfactory performance. The NRC intends that guidance would be 
available to facilitate the review of the requalification examination 
programs that are proposed by facility licensees and that, following 
NRC approval, requalification examination programs would be subject to 
an appropriate change control process.
    For examinations to provide for valid assessments of the knowledge 
and abilities of individuals, the examinations must remain free from 
compromises that could affect their underlying integrity. The NRC is 
proposing to adapt the requirements of Sec.  55.49 in Sec.  53.780(d) 
to require that examinations and related activities remain free from 
any compromise that might affect the integrity of the examination 
process.
    Simulators provide a valuable means of training and evaluating 
plant operators, and the NRC is specifically authorized under the 
Nuclear Waste Policy Act of 1982, as amended (NWPA), section 306 (42 
U.S.C. 10226) to establish regulations for the use of simulators within 
such context. The NRC is proposing to adapt the requirements of Sec.  
55.46 in Sec.  53.780(e) to address the use of simulation facilities 
for training, examinations, and applicant experience requirements, as 
well as to address the maintenance of simulator fidelity. However, the 
proposed requirements of part 53 would not mandate that full scope, 
plant-referenced simulators be used and would allow the use of 
alternative simulation facilities consisting of, for example, partial 
scope simulators or the plant itself, provided that all associated 
requirements can be demonstrated to be met using alternative approaches 
and methods. Additionally, in allowing for the possibility that an 
applicant or licensee might demonstrate compliance with training, 
examination, or experience requirements using the plant itself, the NRC 
is not allowing the initiation of transients on the actual plant. 
Consistent with this, aside from controlled reactivity manipulations 
that are conducted for the purposes of demonstrating compliance with 
experience requirements, actual plant components may not be operated 
for these purposes. Rather, the NRC perspective is that the use of the 
plant for training and examination purposes should be restricted to 
techniques such as walkthroughs, job performance measures, simulated 
tasks, use of augmented reality technology, and similar approaches that 
provide training and examination value while avoiding the operation of 
actual plant components.
    There may be situations in which applicants for operator or senior 
operator licenses have previous training and experience that justifies 
waiving some, or all, of the initial examination requirements. The NRC 
is proposing to adapt the requirements of Sec.  55.47 in Sec.  
53.780(f) to allow for consideration of requests for waivers of 
examinations requirements. In contrast with the part 55 requirements, 
the NRC proposes to locate certain details associated with such waiver 
requests within guidance documentation in lieu of placement within the 
rule itself.
    For licensed operators and senior operators to perform their 
assigned duties safely and reliably, it is essential that they perform 
those duties frequently enough so as to maintain a sufficient degree of 
proficiency. The NRC is proposing to adapt the requirements of Sec.  
55.53(e) and (f) in Sec.  53.780(g) to require that specifically 
licensed operators and senior operators maintain proficiency and, if 
proficiency is not maintained, regain proficiency prior to resuming 
licensed duties. However, in recognition of the fact that varying 
concepts of operations are possible for advanced reactor facilities, 
the NRC is proposing, in contrast with the requirements of part 55, to 
allow facility licensees to establish their own programs for operator 
proficiency, subject to NRC approval.
    As the holders of specific licenses, licensed operators and senior 
operators must be subject to license conditions on an individual basis 
to ensure that the basis upon which the licenses were issued remains 
valid. The NRC is proposing to adapt the requirements of Sec.  55.53 in 
Sec.  53.785 to require appropriate conditions of licenses for 
specifically licensed operators and senior operators. However, in 
contrast with the requirements of Sec.  55.53(e) and (f), the NRC is 
proposing to allow certain aspects of operator proficiency to be 
addressed by an NRC-approved facility proficiency program.
    Licenses for specifically licensed operators and senior operators 
are issued by the NRC and must remain subject to modification or 
revocation. The NRC is proposing to adapt the requirements of 
Sec. Sec.  55.51 and 55.61 in Sec.  53.790 to address the issuance, 
modification, and revocation of licenses issued to specifically 
licensed operators and senior operators.
    The licenses issued to specifically licensed operators and senior 
operators are valid for a period of six years, after which they expire, 
unless otherwise renewed. The NRC is proposing to adapt the 
requirements of Sec. Sec.  55.55 and 55.57 in Sec.  53.795 to address 
the expiration and renewal of licenses issued to specifically licensed 
operators and senior operators.
    In developing this proposed rule, the NRC has discussed with 
stakeholders the considerations that might justify the omission of the 
specifically licensed operators and senior operators. However, even for 
an inherently safe reactor with autonomous operation features, certain 
important administrative functions (e.g., compliance with TS, 
operability determinations, NRC notifications, emergency declarations, 
risk assessment, maintenance oversight, and radiological release limit 
compliance) would still need to be accomplished by

[[Page 86939]]

appropriately qualified and authorized individuals. Additionally, the 
NRC recognized that manual manipulations of facility reactivity 
controls must only be performed by individuals who have been 
appropriately licensed by the Commission. The NRC therefore proposes 
under Sec.  53.800 to establish a new class of facility (defined as a 
self-reliant-mitigation facility), according to the criteria contained 
in Sec.  53.800 for part 53. These facilities would employ GLROs rather 
than specifically licensed operators and senior operators. The GLRO 
regulations offer enhanced flexibilities and targeted relaxations in a 
manner that is commensurate with the modified role of such operators to 
ensure the safe operation of the associated facilities. In contrast, 
those facilities not meeting the criteria of Sec.  53.800 would instead 
be considered interaction-dependent-mitigation facilities and would 
require staffing by specifically licensed operators and senior 
operators. The terminology used to designate these facility types 
reflects differences in how operators are anticipated to need to 
interact with their plant systems in mitigating events and achieving 
safe outcomes; such systems may either need operators to interact with 
them in some manner (i.e., be interaction-dependent) or may instead be 
able to rely fully upon their own capabilities independent of operator 
interaction (i.e., be self-reliant).
    Generally licensed reactor operators would differ from specifically 
licensed operators because the latter would be directly and 
independently evaluated by the NRC as part of their licensing process. 
This direct and independent evaluation remains appropriate when 
operators may reasonably be expected to exert a significant influence 
on public health and safety outcomes. Therefore, a key determinant as 
to whether generally licensed reactor operators can be utilized in 
facility staffing is the assessment of the operator's role in 
maintaining and fulfilling safety functions at the facility, such as 
through the performance of credited actions for the mitigation of plant 
events.
    The criteria proposed in Sec.  53.800 would designate self-reliant-
mitigation facilities. These criteria are derived from the following 
set of considerations:
     no human action needed to satisfy radiological consequence 
criteria;
     no human action needed to address LBEs;
     safety functions not allocated to human action;
     reliance upon robust and highly reliable safety features; 
and
     adequate defense in depth achieved without reliance on 
human action.
    It should be noted that those facilities not meeting the criteria 
proposed in Sec.  53.800 would instead be classified as interaction-
dependent-mitigation facilities and would require staffing by 
specifically licensed operators and senior operators instead.
    Generally licensed reactor operators would perform duties under the 
provisions of a general license that would be effective without the 
filing of an application with the Commission or the issuance of 
licensing documents to a particular person. The NRC proposes 
requirements for the general licensing process for GLROs under 
Sec. Sec.  53.805 through 53.820. The requirements for GLROs would 
parallel those for senior operators in regard to their comparable 
administrative responsibilities. Nonetheless, the requirements for 
GLROs would be relaxed and incorporate greater flexibilities compared 
to the requirements for specifically licensed operators in a manner 
that is consistent with the GLRO's role in safety at self-reliant-
mitigation facilities.
    In order to use GLROs in lieu of specifically licensed operators 
and senior operators, a OL/COL applicant would need to demonstrate that 
its proposed facility is a self-reliant-mitigation facility, i.e., that 
it will comply with the following requirements on an ongoing basis: 
maintaining GLRO qualifications for the performance of important 
functions and tasks; incorporating relevant programmatic controls into 
TS; administering the related programs for training, examination, and 
proficiency; and ensuring that the relevant provisions of parts 26 and 
73 are met. Additionally, to provide for an accurate accounting of what 
individuals are licensed under the general license, facility licensees 
would be required to report the identities of all generally licensed 
reactor operators to the NRC on an annual basis. Furthermore, a 
facility licensee must ensure that the facility design and performance 
continue to meet the technological criteria to be classified as a self-
reliant-mitigation facility (i.e., the criteria of Sec.  53.800) on a 
continual basis during the operating phase, as the relaxations afforded 
to such facilities in the areas of operator licensing, staffing, and 
HFE would be predicated on this assumption. The NRC therefore proposes 
under Sec.  53.805 to establish requirements for facility licensees 
that address issues such as these. Finally, the failure of a self-
reliant-mitigation facility to subsequently meet the criteria of Sec.  
53.800 after the issuance of an OL or COL would constitute a reportable 
event (i.e., an unanalyzed condition that significantly degrades plant 
safety) under the provisions of Sec.  53.1630.
    The NRC proposes the general license for GLROs under Sec.  53.810. 
GLROs would be licensed as a class of individuals under the provision 
of Sec.  53.810(a) and would be subject to the conditions specified in 
Sec.  53.810(b) through (g). Portions of these conditions are adapted 
from Sec.  55.53 and from those conditions currently included in the 
licenses issued to specifically licensed operators and senior 
operators. The NRC would retain the ability to suspend or prohibit 
individuals from operating under the general license should such action 
be warranted.
    The NRC proposes overall programmatic requirements for GLRO 
training, examination, and proficiency under Sec.  53.815. In general, 
these proposed requirements are adapted from those of part 55 and 
parallel those also proposed for specifically licensed senior operators 
in Sec.  53.780. These requirements include increased flexibilities and 
several targeted relaxations that reflect the limited role of GLROs in 
facility safety. The requirements proposed under Sec.  53.815 cover, in 
part, the initial training, initial examination, continuing training, 
requalification examination, and proficiency of GLROs. Section 53.805 
would require the facility licensee to develop, implement, and maintain 
these programs. Section 53.810, in turn, would prescribe that the 
requirements of Sec.  53.805 would need to be met as a requirement of 
the general license. The implication of this structure is that the 
facility licensee would need to implement these programs for training, 
examination, and proficiency, and GLROs would need to participate in 
these programs to demonstrate compliance with the requirements of the 
general license.
    The initial training process provides GLROs with the knowledge and 
abilities needed to fulfill assigned duties as GLROs. The use of a SAT 
serves to ensure that the training program is based upon job 
requirements in a manner that can be adapted to account for differences 
in plant technology and concepts of operations. The NRC is proposing 
under Sec.  53.815(b) to require facility licensees to implement a SAT-
based training program for the initial training of GLROs that is 
adequate to ensure that they have the necessary knowledge, skills, and 
abilities to perform their duties. The NRC further proposes that such 
programs would be subject to NRC approval, oversight, and appropriate 
change control processes. The training program must ensure that

[[Page 86940]]

GLROs maintain the necessary knowledge, skills, and abilities.
    Examinations provide a means of assessing that individuals have 
achieved a degree of knowledge and ability that will be sufficient to 
enable them to carry out assigned duties as GLROs in a manner that is 
both safe and reliable. The NRC proposes to adapt the requirements of 
Sec. Sec.  55.40, 55.41, 55.43, and 55.45 in Sec.  53.815(b) to require 
that facility licensees establish and implement an initial examination 
program. A key difference from the comparable requirements of part 55 
would be that facility licensees would be afforded the flexibility to 
propose, subject to NRC approval, the examination methods and criteria 
to be used in assessing satisfactory individual performance. Such 
examination programs (including those used within the scope of 
continuing training) would need to provide for acceptable levels of 
both test validity and test reliability in order to be considered 
acceptable. The NRC intends that staff guidance would be available to 
facilitate the review of initial examination programs that are proposed 
by facility licensees and that approved initial examination programs 
would be subject to an appropriate change control process. In contrast 
with both the requirements of part 55 and the proposed requirements of 
Sec.  53.780, the NRC does not intend to administer or evaluate these 
initial examinations. However, the examination processes themselves 
will continue to be subject to ongoing NRC oversight. Irrespective of 
the provided flexibilities in examination format and structure, topics 
from the following general categories of knowledge and abilities should 
be sampled in such examinations:

 Reactor Theory, Thermodynamics, and Chemical Interactions
 Plant Systems and Components
 Reactivity Management and Manipulations
 Radiation Control and Safety
 Emergency, Abnormal, and Normal Operations
 Administrative Requirements and Conditions of the Facility 
License

    Continuing training programs provide the ongoing training and 
examination of GLROs to ensure that they maintain the knowledge and 
abilities needed to support the safe and reliable performance of job 
duties following the completion of an initial training and examination 
program. The NRC is proposing to adapt the requirements of Sec.  55.59 
in Sec.  53.815(b) to require that facility licensees implement both a 
SAT-based continuing training program and a requalification examination 
program. However, a notable difference from the examinations required 
under part 55 would be that distinct annual operating test and biennial 
written examination components would not be mandated. The facility 
licensee would instead propose examination methods and criteria to be 
used in assessing satisfactory performance. Furthermore, unlike the 
comparable requirements of part 55 and those proposed for specifically 
licensed operators and senior operators, a biennial periodicity for 
requalification examinations would not be prescribed. However, adequate 
justification for the proposed periodicity of requalification 
examinations would be required. The NRC intends that staff guidance 
would be available to facilitate the review of the requalification 
examination programs that are proposed by facility licensees. Approved 
requalification examination programs would be subject to an appropriate 
change control process.
    For examinations to provide for valid assessments of the knowledge 
and abilities of individuals, the examinations must remain free from 
compromises that could affect their underlying integrity. The NRC is 
proposing to adapt the requirements of Sec.  55.49 in Sec.  53.815(d) 
to require that examinations and related activities remain free from 
any compromise that might affect the integrity of the examination 
process.
    Simulators provide a valuable means of training and evaluating 
plant operators and the NRC is specifically authorized under the NWPA, 
section 306 (42 U.S.C. 10226) to establish regulations for the use of 
simulators within such context. The NRC is proposing to adapt the 
requirements of Sec.  55.46 in Sec.  53.815(e) to address the use of 
simulation facilities for training and examinations, and experience 
requirements, as well as to address the maintenance of simulator 
fidelity. The use of full scope, plant-referenced simulators would not 
be mandated. The potential use of alternative simulation facilities 
consisting of, for example, partial scope simulators or the plant 
itself, would be allowed provided that all associated requirements 
could be demonstrated to be met using alternative approaches and 
methods. Additionally, in allowing for the possibility that an 
applicant or licensee might demonstrate compliance with training and 
examination requirements using the plant itself, the NRC is not 
allowing the initiation of transients on the actual plant. Consistent 
with this, aside from controlled reactivity manipulations that are 
conducted for the purposes of demonstrating compliance with experience 
requirements, actual plant components may not be operated for these 
purposes. Rather, the use of the plant for training and examination 
purposes should be restricted to techniques such as walkthroughs, job 
performance measures, simulated tasks, use of augmented reality 
technology, and similar approaches that provide training and 
examination value while avoiding the operation of actual plant 
components.
    There may be situations in which GLROs have previous training and 
experience that justifies waiving some, or all, of the initial 
examination. Therefore, the NRC is proposing under Sec.  53.815(f) to 
allow facility licensees to waive some, or all, portions of initial 
examinations provided that such waivers are consistent with a program 
that has been approved by the NRC.
    For GLROs to safely and reliably perform their assigned duties, it 
is essential that they perform those duties frequently enough so as to 
maintain a sufficient degree of proficiency. However, the NRC 
recognizes that facilities that utilize GLROs may have concepts of 
operation that warrant unique proficiency considerations. Therefore, 
the NRC is proposing in Sec.  53.815(g) to require that facility 
licensees develop, implement, and maintain programs to maintain and 
reestablish, if needed, the proficiency of GLROs. This could occur, for 
example, if an individual's extended absence from watch standing has 
rendered proficiency requirements unmet.
    The general license should remain in effect for an individual only 
while that individual remains employed in a position that may call for 
the individual to manipulate the reactivity controls of the facility. 
The NRC proposes under Sec.  53.820 to require that the general license 
would cease to be applicable on an individual basis when an 
individual's employment status becomes such that this is no longer the 
case. However, the NRC recognizes that for some types of self-reliant-
mitigation facilities, very long periods may elapse between 
circumstances that necessitate manual manipulation of reactivity 
controls. Therefore, the general license remains in effect for an 
individual as long as the individual's current position could 
potentially require that individual to manipulate reactivity controls 
at some point within the course of the individual's assigned job 
duties.
    The NWPA, section 306 (42 U.S.C. 10226) authorizes and directs the 
NRC to, in part, issue regulations and guidance that address the 
training and

[[Page 86941]]

qualifications of civilian nuclear power plant operators, supervisors, 
technicians, and other appropriate operating personnel. The NRC 
implements this in part 50 through the requirements of Sec.  50.120, 
``Training and qualification of nuclear power plant personnel.'' The 
NRC is proposing under Sec.  53.830 to adapt, with modifications, the 
requirements of Sec.  50.120 for use in part 53 to provide more 
flexible personnel training and qualification requirements than those 
in Sec.  50.120 and better reflect diverse concepts of operations.
    The NRC recognizes that the categories of nuclear power plant 
personnel in Sec.  50.120 may not be needed for the diverse concepts of 
operations, staffing models, and non-traditional personnel roles and 
responsibilities anticipated under proposed part 53; conversely, and 
for the same reasons, additional categories of plant personnel may need 
to be covered by part 53. The NRC also recognizes that the timeframe 
prescribed in Sec.  50.120 for the establishment of training programs 
may not be aligned with the schedules associated with the startup of 
certain types of commercial nuclear plant facilities. However, the NRC 
also recognizes that the SAT-based training required under Sec.  50.120 
remains an appropriate means by which training programs should continue 
to be developed and implemented. Therefore, the approach taken by the 
NRC in addressing the training of certain plant staff under the 
proposed part 53 reflects greater flexibilities in personnel categories 
and programmatic timeframes, while still retaining the requirement that 
such training programs be based on SAT.
    The NRC is proposing under Sec.  53.830 to require SAT-based 
training programs with the timeframe for when such programs are 
required being based upon when the associated personnel are needed to 
support facility-specific needs. The training programs would cover the 
training and qualification of plant personnel in the general categories 
of supervisors, technicians, and other appropriate operating personnel. 
The licensee would not be required to seek NRC approval of a training 
program prior to usage. However, the licensee is required to 
accommodate NRC inspection of the training program. The NRC intends to 
develop guidance to facilitate the inspection of these training 
programs but does not intend for such guidance to preclude the 
potential for the training programs to be maintained by a separate, 
NRC-approved accreditation process.
    The proposed Sec.  53.845 would require programs to be developed, 
implemented, and maintained to help ensure that design features and 
human actions have the capabilities and reliabilities necessary to 
demonstrate compliance with the safety criteria in subpart B throughout 
the operating life of each commercial nuclear plant. The proposed 
programmatic requirements in subpart F would also address areas such as 
radiation protection needed to control routine effluents during normal 
operations. The proposed Sec. Sec.  53.850 through 53.910 would require 
programs to support specific activities needed to ensure the prevention 
or mitigation of unplanned events or to support normal operations for 
any reactor design. However, each holder of an OL or COL would be 
required to assess whether additional programs are needed for the 
specific reactor design and location of the commercial nuclear plant. 
Licensees would be able to combine, separate, and otherwise organize 
programs and related documents as appropriate for the technologies and 
organizations associated with the commercial nuclear plant.
    Proposed Sec.  53.850 would require a radiation protection program 
associated with the requirements in subparts B and C for public doses 
resulting from normal operations and the protection of plant workers. 
The proposed requirements related to doses from normal operations, 
including routine effluents, would be similar to those specified in 
Sec.  50.36a, ``Technical specifications on effluents from nuclear 
power reactors,'' and related requirements in standard TS for offsite 
dose calculation manuals. While the proposed section would include 
requirements that are technically and programmatically similar to part 
50, proposed Sec.  53.850 would not include a requirement for effluent-
related TS as is required in Sec.  50.36a. A proposed requirement 
similar to that found in the administrative controls section of TS for 
operating reactors licensed under parts 50 and 52 would be included for 
programmatic controls of solid wastes to complement the design 
requirements in proposed Sec.  53.425.
    Proposed Sec.  53.855 would require an emergency response plan that 
demonstrates compliance with the requirements in appendix E to part 50 
and Sec.  50.47(b) or Sec.  50.160. The regulations in Sec.  50.47 
stating that the NRC will not issue certain licenses unless it finds 
that there is reasonable assurance that adequate protective measures 
can and will be taken to protect public health and safety in the event 
of a radiological emergency apply equally to applications under part 53 
complying with the applicable standards set forth in either Sec.  
50.160 or the requirements in appendix E to part 50 and Sec.  50.47(b).
    In its 2008 Advanced Reactor Policy Statement, the Commission 
stated their expectation that ``the safety features of advanced reactor 
designs will be complemented by the operational program for Emergency 
Planning (EP). This EP operational program, in turn, must be 
demonstrated by inspections, tests, analyses, and acceptance criteria 
to ensure effective implementation of established measures.'' 
Consistent with this policy statement, emergency plans and emergency 
planning zones are not safety features in the design. In SECY-97-020, 
``Results of Evaluation of Emergency Planning for Evolutionary and 
Advanced Reactors,'' dated January 27, 1997, the staff indicated that 
the rationale upon which EP for current reactor designs is based, that 
is, potential consequences from a spectrum of accidents, is appropriate 
for use as the basis for EP for evolutionary and passive advanced LWR 
designs and is consistent with the Commission's defense-in-depth safety 
philosophy. Also, in its Safety Goals Policy Statement the Commission 
stated that: ``A defense-in-depth approach has been mandated in order 
to prevent accidents from happening and to mitigate their consequences. 
Siting in less populated areas is emphasized. Furthermore, emergency 
response capabilities are mandated to provide additional defense-in-
depth protection to the surrounding population.'' Consistent with this 
policy statement, proposed Sec.  53.855 contributes an additional 
independent layer of defense in depth for commercial nuclear plants. 
Therefore, the emergency plans and emergency planning zones under 
proposed Sec.  53.855 are not used to demonstrate compliance with 
subpart B and subpart C of this part. Rather, compliance with the 
requirements in proposed Sec.  53.855 would provide reasonable 
assurance that adequate protective measures can and will be taken to 
protect public health and safety in the event of a radiological 
emergency.
    Proposed Sec.  53.860 would identify the applicable regulations for 
part 53 applicants related to the programs for physical security, 
cybersecurity, FFD, AA, and information security. These programs are 
discussed in more detail in section V, ``Changes to Other Parts of 10 
CFR,'' of this document.
    Proposed Sec.  53.860(a) would establish the physical protection 
program and present a graded approach to physical protection 
requirements. If a licensee can meet the proposed criterion in

[[Page 86942]]

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

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

    Proposed Sec.  53.860(b) would require licensees to establish, 
implement, and maintain an FFD program under part 26. Section 53.860(c) 
would require licensees to establish, implement, and maintain an AA 
program in accordance with either Sec.  73.56 or proposed Sec.  73.120, 
as appropriate. Section 53.860(d) would require licensees to establish, 
implement, and maintain a cybersecurity program in accordance with 
either Sec.  73.54 or proposed Sec.  73.110. Section 53.860(e) would 
require licensees to establish, implement, and maintain an information 
protection system that complies with the requirements of Sec. Sec.  
73.21, 73.22, and 73.23, as applicable.
    Proposed Sec.  53.865 would establish requirements for quality 
assurance and refer to appendix B of part 50 for the part 53 
requirements for SR design features. Proposed requirements related to 
evaluating and reporting changes to the quality assurance program would 
be included in proposed subpart I and would be equivalent to those 
found in Sec.  50.54.
    The proposed Sec.  53.870 would require licensees to actively 
assess possible degradation of SSCs from the effects of aging, fatigue, 
and environmental conditions. The proposed inclusion of requirements 
related to designing and monitoring for possible degradation mechanisms 
reflects important lessons learned from the history of LWRs and the 
likely introduction of new design features and materials in future 
commercial nuclear plants. The allowable combinations of design 
features, operating experience, testing, and monitoring during 
operations would support performance-based approaches to the initial 
licensing of new technologies. The proposed performance-based approach 
to integrity assessment programs would also allow for the subsequent 
consideration of operating experience and appropriate corrective 
actions or allowable relaxations for ensuring that design features 
comply with the proposed functional design criteria of Sec. Sec.  
53.410 and 53.420. The proposed program would be based upon a 
comprehensive and integrated evaluation of the aging and other 
degradation mechanisms applicable to the design; identification of the 
affected SSCs; the allowances provided in the design of the SSCs for 
degradation; and schedules and procedures for determining if and at 
what rate degradation is occurring, as well as its cause. Risk insights 
could be used to prioritize the monitoring, evaluation, and management 
of degradation based upon the importance of the SSC to safety and the 
time frame for when the effects of degradation could be of concern.
    Proposed Sec.  53.875 would establish requirements for a fire 
protection program supporting operations similar to Sec.  50.48. The 
proposed fire protection program during operations would work in 
concert with specific fire protection requirements proposed in subpart 
C for design and analyses and in proposed subpart E for construction 
and manufacturing.
    Proposed Sec.  53.880 would establish requirements for an inservice 
inspection (ISI) and inservice testing (IST) program, which are 
historically important activities conducted in accordance with ASME 
codes and regulations in Sec.  50.55a. While the proposed part 53 would 
not incorporate specific consensus codes and standards into the 
regulations, Sec.  53.880 allows for the use of generally accepted 
codes and standards. The proposed requirement for an ISI and IST 
program would reinforce the need to develop monitoring programs to be 
conducted during a plant's operations phase to complement the design 
process and address inherent uncertainties. The NRC encourages the 
continued use of consensus codes and standards supporting design, 
testing, and inspections to support integrated and performance-based 
approaches in demonstrating compliance with the proposed requirements 
in part 53.
    Proposed Sec.  53.910 would establish requirements for developing, 
implementing, and maintaining procedures (e.g., operations and 
emergency operating procedures) and guidelines (e.g., accident 
management guidelines). The programmatic requirements for many of the 
procedures listed in this proposed section would be similar to the 
requirements found in the administrative controls section of TS for 
plants licensed under parts 50 and 52. The proposed inclusion, where 
appropriate, of accident management guidelines in these requirements is 
intended to ensure that an integrated set of procedures and guidelines 
would be established by licensees to ensure command and control across 
the spectrum of possible event sequences. The proposed required 
procedures would also include those needed to complement the design 
requirements in proposed Sec.  53.440(m) related to criticality alarms 
and the equivalent of the procedures required in Sec.  50.54(hh) to 
address notifications of potential aircraft threats.

Subpart G--Decommissioning Requirements

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

[[Page 86943]]

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

Subpart H--Licenses, Certifications, and Approvals

    Proposed subpart H would provide requirements related to 
applications under part 53 for NRC licenses, certifications, or 
approvals for commercial nuclear plants.
    Proposed subpart H would specify requirements applicable to all 
part 53 applications as well as requirements specific to part 53 
applications for LWAs, ESPs, standard design approvals, standard DCs, 
MLs, CPs, OLs, and COLs. Proposed subpart H would be equivalent to and 
include all existing licensing, certification, and approval processes 
currently covered under parts 50 and 52, with the exception of the 
process for early review of site suitability issues. Interactions with 
external stakeholders during the development of the proposed rule did 
not identify significant interest in or need for including the process 
for early review of site suitability issues in part 53.
    Much of the proposed subpart H regulatory text is identical to the 
corresponding language in parts 50 and 52, with minor changes to 
account for cross references in part 53, to make language technology 
neutral, or to reflect the unique analytical approach in part 53. In 
these instances, this preamble discussion will describe the language as 
``equivalent'' to the existing corresponding requirement in part 50 or 
part 52 and will describe any deviations, where applicable.
    Because part 53 carries over the majority of the licensing options 
from parts 50 and 52, there are several sections in proposed subpart H 
that are similar to existing regulations in parts 50 and 52. Proposed 
Sec.  53.1100 would address filing of applications for licenses, 
certifications, or approvals under oath or affirmation and is 
equivalent to Sec.  50.30. The proposed Sec.  53.1100 does not include 
the current requirement in Sec.  50.30(a)(2) that the applicant 
maintain the capability to generate additional copies, because it is 
unnecessary in the age of electronic submissions. In addition, the 
existing requirement on applications for OLs in Sec.  50.30(d) is 
included in proposed Sec.  53.1124(g)(2), ``Relationship between 
sections,'' covering OLs, rather than in proposed Sec.  53.1100.
    Proposed Sec.  53.1101 would lay out activities requiring an NRC 
license and is equivalent to Sec.  50.10(b). Proposed Sec.  53.1103 
would address combining applications and is equivalent to Sec. Sec.  
50.31, 50.52, and 52.8. Proposed Sec.  53.1103(b) would continue the 
Commission's practice of combining multiple authorizations for a 
facility under parts 30, 40, 50, 52, and 70 into one license based on 
the Commission's authority under Section 161h. of the Act to combine 
NRC licenses. Proposed Sec.  53.1106 would address elimination of 
repetition and is equivalent to Sec.  50.32.
    Proposed Sec.  53.1109 would provide general information 
requirements for the content of applications submitted to the NRC under 
part 53 and is equivalent to Sec.  50.33, with the exception of Sec.  
50.33(f) on financial qualifications, which is covered in proposed 
subpart J, and Sec.  50.33(h) on earliest and latest dates for 
completion of construction, which is covered in Sec.  53.1306 of this 
subpart. Each application would need to include information to address 
the items in proposed Sec.  53.1109 as cited in the appropriate section 
of this subpart for the application type.
    One change from current requirements can be found in proposed Sec.  
53.1109(i), which is not limited to electricity generation as it is 
currently in part 50. Some prospective NRC applicants are considering 
development of nuclear plants for other commercial ventures, such as 
process heat generation or hydrogen production. In addition, Sec.  
53.1109(j), which requires applications containing classified 
information to separate that information from the unclassified 
information in the application, refers to ``Restricted Data or 
classified National Security Information'' instead of the term used in 
the corresponding provision in Sec.  50.33(j), ``Restricted Data or 
other defense information.'' This change was made to use the defined 
term in part 95 rather than ``defense information'' as used in Sec.  
50.33(j). The usage in Sec.  50.33(j) dates back to the Atomic Energy 
Commission amendment of that section on January 19, 1956 (21 FR 355, 
357) and was not changed with the issuance of part 95 (45 FR 14476; 
March 5, 1980) after the establishment of the NRC and the 1975 
reissuance of the former Atomic Energy Commission regulations. The 
revised terminology also aligns with its usage in Sec.  53.1115.
    Proposed Sec.  53.1112 would address environmental conditions and 
is equivalent to Sec.  50.36b. Proposed Sec.  53.1115 would address 
requirements for agreements limiting access to classified information 
and is equivalent to Sec.  50.37. Proposed Sec.  53.1118 would address 
ineligibility of certain applicants and is equivalent to Sec.  50.38. 
Proposed Sec.  53.1120 would address exceptions and exemptions from 
licensing requirements for Department of Defense and DOE facilities and 
is equivalent to Sec.  50.11. Proposed Sec.  53.1121 would address 
public inspection of applications and is equivalent to Sec.  50.39.
    Proposed Sec.  53.1124 would address the relationship between the 
various licenses, certifications, and approvals

[[Page 86944]]

provided in this subpart, and the requirements are equivalent to a 
number of similar provisions in parts 50 and 52 including Sec. Sec.  
50.10, 52.13, 52.43, 52.73, 52.133, and 52.153. New provisions are 
provided in Sec.  53.1124(c) and (d), that would allow an application 
for either a standard design approval or a standard DC under part 53 to 
reference applicable licensing-basis information that supported 
issuance of an OL or COL under part 53. These provisions are being 
proposed to offer additional flexibility beyond what is currently 
allowed under parts 50 or 52 for an applicant who may wish to license a 
first-of-a-kind reactor for operation prior to seeking generic approval 
or certification of the standard design.
    Proposed Sec.  53.1124(e) would address the limitations that a 
manufactured reactor may only be transported to a site with a COL and 
is equivalent to Sec.  52.153. Proposed Sec.  53.1130 would address 
LWAs and is equivalent to Sec.  50.10.
    Proposed Sec. Sec.  53.1140 through 53.1188 would govern the 
content of ESP applications. Proposed Sec.  53.1140 is equivalent to 
Sec.  52.12. Proposed Sec.  53.1143 would address filing of 
applications and is equivalent to Sec.  52.15. Proposed Sec.  53.1144 
would address general information requirements for the content of 
applications and is equivalent to Sec.  52.16.
    Proposed Sec.  53.1146 would specify requirements for the technical 
contents of applications and is equivalent to Sec.  52.17. Proposed 
Sec.  53.1146(b)(2) provides applicants for ESPs a regulatory option to 
propose major features of the emergency plans or complete and 
integrated emergency plans in accordance with either the requirements 
in Sec.  50.160 of this chapter, or the requirements in appendix E to 
part 50 of this chapter and Sec.  50.47(b) of this chapter, as 
applicable.
    Proposed Sec.  53.1149 would address standards for review of ESP 
applications and administrative review of applications, including 
hearings, and is equivalent to Sec. Sec.  52.18 and 52.21. Proposed 
Sec.  53.1155 would address referral to the ACRS and is equivalent to 
Sec.  52.23. Proposed Sec.  53.1158 would address issuance of ESPs and 
is equivalent to Sec.  52.24. Proposed Sec.  53.1161 would address the 
extent of activities permitted and is equivalent to Sec.  52.25. 
Proposed Sec.  53.1164 would address the duration of an ESP and is 
equivalent to Sec.  52.26. Proposed Sec.  53.1167 would address 
provisions for requesting a LWA after issuance of an ESP and is 
equivalent to Sec.  52.27. Proposed Sec.  53.1170 would address 
transfers of ESPs and is equivalent to Sec.  52.28. Proposed Sec.  
53.1173 would address applications for ESP renewals and is equivalent 
to Sec.  52.29. Proposed Sec.  53.1176 would address criteria for 
renewal of an ESP and is equivalent to Sec.  52.31. Proposed Sec.  
53.1179 would address the duration of an ESP renewal and is equivalent 
to Sec.  52.33. Proposed Sec.  53.1182 would address the use of a site 
for purposes other than those described in the permit and is equivalent 
to Sec.  52.35. Proposed Sec.  53.1188 would address finality of ESP 
determinations and is equivalent to Sec.  52.39.
    Proposed Sec. Sec.  53.1200 through 53.1221 would govern the 
contents of standard design approval applications. Proposed Sec.  
53.1200 is equivalent to Sec.  52.131. Proposed Sec.  53.1203 would 
address filing of applications and is equivalent to Sec.  52.135. 
Proposed Sec.  53.1206 would address general information requirements 
for the content of applications and is equivalent to Sec.  52.136.
    Proposed Sec.  53.1209 would address requirements for the technical 
content of applications and is largely equivalent to Sec.  52.137. In 
proposed Sec.  53.1209(a), the NRC proposes text that expands the 
discussion of ``major portion'' standard design approvals. Additional 
discussion regarding standard design approvals for a major portion of a 
standard design can be found in the NRC's ``A Regulatory Review Roadmap 
for Non-Light Water Reactors,'' which considers the Nuclear Innovation 
Alliance report ``Clarifying `Major Portions' of a Reactor Design in 
Support of a Standard Design Approval.'' Proposed Sec.  53.1209(b) 
outlines the required content of the Final Safety Analysis Report 
(FSAR). Proposed requirements in Sec.  53.1209(b)(2) for portions of 
the application addressing design information state that the 
application must include design information equivalent to that required 
for a standard DC. This reference to the pertinent DC requirements 
(specifically, those in proposed Sec.  53.1239(a)(2) through (27)) is 
an efficiency that would prevent the need to repeat many of the same 
requirements for the content of a standard design approval application.
    Proposed Sec.  53.1210 would address requirements for the content 
of a standard design approval application other than the FSAR. Proposed 
Sec.  53.1210(a) would require the inclusion of a description of 
availability controls that are not included in the FSAR.
    Proposed Sec.  53.1212 would address standards for review of 
applications and is equivalent to Sec.  52.139. Proposed Sec.  53.1215 
would address referral to the ACRS and is equivalent to Sec.  52.141. 
Proposed Sec.  53.1218 would address staff approval of designs and 
duration of design approvals and is equivalent to Sec. Sec.  52.143 and 
52.147. Proposed Sec.  53.1221 would address finality of standard 
design approvals and information requests and is equivalent to Sec.  
52.145 with the exception that it extends such finality to a standard 
approval referenced in a DC application. Standard design approvals 
issued to date under part 52 have been issued during the NRC's review 
of the standard DC application and have relied on the same application 
content. However, a future scenario could arise where the DC 
application is not submitted until after a design approval has been 
granted. The NRC would apply the same finality provisions in this 
situation as in the situation where a standard design approval is 
referenced in a COL application.
    There is no equivalent to proposed Sec.  53.1221(d) in part 52 for 
standard design approvals. This provision would state that the 
Commission will require, before granting a CP, COL, OL, or ML which 
references a standard design approval, that engineering documents be 
completed and available for audit. A similar provision is included in 
part 52 in relation to a standard DC; and the NRC would require that 
design and analysis information needed for the Commission to make its 
safety determination be complete and available for any application the 
NRC is reviewing. Making this explicit provides increased clarity to 
future standard design approval applicants under part 53.
    Proposed Sec. Sec.  53.1230 through 53.1263 would address standard 
DCs. Proposed Sec.  53.1230 would address general provisions for 
standard DCs and is equivalent to Sec.  52.41. Proposed Sec.  53.1233 
would address filing of applications and is equivalent to Sec.  52.45. 
Proposed Sec.  53.1236 would address general information requirements 
for the content of applications and is equivalent to Sec.  52.46. 
Proposed Sec.  53.1239 would address requirements for the technical 
content of applications and is equivalent to Sec.  52.47(a). The 
requirements in proposed Sec.  53.1239 have been modified from the 
analogous requirements in Sec.  52.47(a) to align with the technical 
requirements in proposed part 53.
    Proposed Sec.  53.1241 would address requirements for the content 
of a standard DC application other than the FSAR and is equivalent to 
Sec.  52.47(b) and (d).

[[Page 86945]]

    Proposed Sec.  53.1242 would address review of applications and is 
equivalent to Sec. Sec.  52.48 and 52.51. Proposed Sec.  53.1242(c) 
would include a provision that would allow a DC applicant to reference 
applicable licensing-basis information for an OL or COL issued under 
part 53. As explained previously, this provision is being proposed to 
explicitly allow flexibility for an applicant who may wish to license a 
first-of-a-kind reactor for operation prior to seeking certification of 
the generic reactor design. For NRC findings on a reactor design in an 
OL or COL proceeding, this proposal would provide finality in a 
subsequent DC application that references information on the OL or COL 
proceeding's docket. This finality accorded to the OL or COL findings 
would bind the NRC staff and the ACRS but would not bind members of the 
public or the Commission. (To the extent an Atomic Safety and Licensing 
Board (ASLB) might have a role in a DC rulemaking, the OL or COL 
findings would not bind the ASLB either.) Specifically, members of the 
public would have the opportunity to comment on a proposed DC rule 
under well-established NRC practice. The rationale for binding the NRC 
staff and ACRS is similar to the rationale for a COL applicant 
referencing a standard design approval under part 52.
    Proposed Sec.  53.1245 would address referral to the ACRS and is 
equivalent to Sec.  52.53. Proposed Sec.  53.1248 would address 
issuance of standard DCs and is equivalent to Sec.  52.54. Proposed 
Sec.  53.1251 would address duration of certifications and is 
equivalent to Sec.  52.55(c). Proposed Sec.  53.1254 would address 
application for renewal and is equivalent to Sec.  52.57. Proposed 
Sec.  53.1257 would address criteria for renewal and is equivalent to 
Sec.  52.59. Proposed Sec.  53.1260 would address duration of renewals 
and is equivalent to Sec.  52.61. Proposed Sec.  53.1263 would address 
finality of standard DCs and is equivalent to Sec.  52.63.
    Proposed Sec. Sec.  53.1270 through 53.1291 would address MLs 
covering manufacturing activities at one or more licensee facilities. 
Proposed Sec.  53.1270 would address the scope of these sections and is 
equivalent to Sec.  52.151.
    Proposed Sec.  53.1273 would address filing of applications for an 
ML and is equivalent to Sec.  52.155(a).
    Proposed Sec.  53.1276 would address general information 
requirements for the content of ML applications and is equivalent to 
Sec.  52.156, with one exception. Proposed Sec.  53.1276 would require 
each application for an ML to also include the information required by 
Sec.  53.1109(e). This information includes the type of license applied 
for, the use to which the facility will be put, the period of time for 
which the license is sought, and a list of other licenses, except 
operator's licenses, issued or applied for in connection with the 
proposed facility to address the potential variations in how MLs might 
be formulated under the proposed part 53.
    Proposed Sec.  53.1279 would address requirements for the technical 
content of applications for MLs to be included in the FSAR and is 
equivalent to Sec.  52.157. In addition, the requirements in proposed 
Sec.  53.1279(a) and (b) have been modified from the analogous 
requirements in Sec.  52.157 to align with the technical requirements 
in proposed part 53. Proposed Sec.  53.1279(a)(2) outlines the required 
content of the application addressing design information and states 
that the application must include design information equivalent to that 
required for a standard DC. This reference to the pertinent DC 
requirements is an efficiency that would prevent the need to repeat the 
same requirements for the content of an ML application.
    Proposed Sec.  53.1279(c) would provide application requirements 
related to the deployment of the completed manufactured reactor. 
Proposed Sec.  53.1279(c)(1) would require inclusion of information 
related to the procedures governing the preparation of the manufactured 
reactor for shipping to the site where it is to be operated, the 
conduct of shipping, and the verification of the condition of the 
shipped items upon receipt at the site. Proposed Sec.  53.1279(c)(2) 
would require that the application include information on the 
interaction of the design, manufacture, and installation of a 
manufactured reactor within the applicant's organization and the manner 
by which the applicant will ensure close integration between the 
designer, contractors, and any licensee of a facility in which the 
manufactured reactor is to be installed. Finally, proposed Sec.  
53.1279(c)(3) would require that the application include a description 
of the measures used for the control of interfaces between the holder 
of the ML and the holder of the COL for the commercial nuclear plant at 
which the manufactured reactor is to be installed. This information is 
necessary for the NRC to determine whether the applicant would have 
appropriate controls in place to ensure coordination between parties 
involved in the design, manufacture, and eventual operation of any 
reactor manufactured under an ML.
    Proposed Sec.  53.1279(d) would include additional requirements for 
application content for applicants seeking an ML for manufactured 
reactors that will be fueled at the factory under a 10 CFR part 70 
license, consistent with the requirements in Sec.  53.620(d). These 
provisions would require the application to include information related 
to loading fuel and the required independent physical mechanisms to 
prevent criticality and to otherwise provide assurance that the fueled 
manufactured reactor can be successfully transported, installed, and 
operated at a site for which the Commission has issued a COL that 
authorizes construction and operation of a commercial nuclear plant 
using the manufactured reactor.
    Proposed Sec.  53.1282 would provide requirements for other 
application content for MLs and is equivalent to Sec.  52.158. Proposed 
Sec.  53.1282(a)(1) would provide requirements to include in the ML 
application the ITAAC within the scope of the ML that the COL holder 
referencing the ML must satisfy. Proposed Sec.  53.1282(a)(2) would 
require that the ITAAC from a referenced standard design apply to the 
portions of the ML design within the scope of the referenced standard 
design. Proposed Sec.  53.1282(a)(3) would state that the COL 
application may include a notification that required referenced 
standard DC ITAAC have been satisfied at the manufacturing facility.
    Proposed Sec.  53.1282(b) would require an ML application to 
include an environmental report and, consistent with existing 
requirements, proposed Sec.  53.1282(b)(2) would note that if the ML 
application references a standard DC, the environmental report need not 
contain a discussion of severe accident mitigation design alternatives 
for the manufactured reactor as used in a commercial nuclear plant.
    Proposed Sec.  53.1285 would provide standards for review of 
applications and administrative review of applications for MLs, 
including hearings, and is equivalent to Sec. Sec.  52.159 and 52.163.
    Proposed Sec.  53.1286 would address referral of applications to 
the ACRS and is equivalent to Sec.  52.165. Proposed Sec.  53.1287 
would address issuance of an ML and is equivalent to Sec.  52.167.
    Proposed Sec.  53.1288 would address finality of MLs and is 
equivalent to Sec.  52.171. Proposed Sec.  53.1291 would address the 
duration of MLs and is equivalent to Sec.  52.173. Proposed Sec.  
53.1293 would address the transfer of MLs and is equivalent to Sec.  
52.175. Proposed Sec.  53.1295 would address the renewal of MLs and is 
equivalent to Sec. Sec.  52.177, 52.179 and 52.181, with a minor 
exception. Proposed Sec.  53.1295(a)(3) would state that an ML

[[Page 86946]]

for which a timely application for renewal has been filed remains in 
effect until the Commission has made a final determination on the 
renewal application, provided, however, that the holder of an ML may 
not begin manufacture of a manufactured reactor less than six months 
before the expiration of the license. The proposed 6-month time frame 
for this provision is changed from the 3-year period in the equivalent 
provision in part 52 because future reactor applicants may present 
smaller, simpler designs, to include micro-reactor designs, in ML 
applications than those that were envisioned when the existing 
requirements were written. A 6-month time frame for this provision 
would provide greater flexibility for ML holders related to 
manufactured reactors being produced when the ML expires.
    Proposed Sec. Sec.  53.1300 through 53.1348 would address licensing 
requirements for CPs. Proposed Sec.  53.1300 would set out general 
requirements for CPs and is equivalent to Sec.  50.23. Proposed Sec.  
53.1306 would address the general information requirements for the 
content of applications for CPs and is equivalent to Sec.  50.33(f) and 
(h).
    Proposed Sec.  53.1309 would address requirements for the technical 
content of applications for CPs and includes the requirement to submit 
a Preliminary Safety Analysis Report (PSAR) that describes the facility 
and presents a preliminary safety analysis of the facility as a whole. 
This is in contrast to an OL application which is required to include 
an FSAR that describes the facility and presents a final safety 
analysis of the facility as a whole. Proposed Sec.  53.1309 is 
equivalent to Sec.  52.17(a)(1)(iv) through (a)(1)(x) and 52.17(b), 
with two exceptions. First, proposed Sec.  53.1309 would replace the 
analysis of the dose criteria required by Sec.  52.17(a)(1)(ix) with 
analysis to demonstrate compliance with the safety criteria defined in 
Sec. Sec.  53.210 and 53.220. Second, proposed Sec.  53.1309(a)(2) 
would add a requirement for a CP application to include several 
categories of detailed design information, although Sec.  
53.1309(a)(2)(ii) would allow certain relaxations of this requirement 
in view of aspects of a design that may not yet be fully developed. 
Section 53.1309 would reference the requirements for the content of an 
ESP application to address application requirements related to siting 
and would reference the requirements for the content of a DC 
application to address application requirements related to design of 
the commercial nuclear plant. Proposed Sec.  53.1309(a)(2)(ii) would 
address the treatment of preliminary design information and notes that 
information provided in the application may include some aspects of the 
design that are not fully developed. This provision would require that 
the completed design, including any changes during construction, be 
described in the FSAR in an application for an OL. This would include 
the requirement for a description of the PRA required by Sec.  
53.450(a) and its results. Probabilistic risk assessments developed for 
commercial nuclear plants prior to construction would be based on the 
design and other information available at the time of the CP 
application. PRAs performed in early design stages or prior to 
construction may be inherently less detailed and may include projected 
information that will be subsequently verified or revised when the 
plant is built. Proposed Sec.  53.1309(a)(4) would address preliminary 
description of the plans for coping with emergencies.
    Proposed Sec.  53.1312 would address other application content for 
CPs. Proposed Sec.  53.1312(a)(1) is equivalent to Sec.  52.80(b) but 
is adapted for a CP application. Proposed Sec.  53.1312(a)(2) is 
equivalent to Sec.  52.80(c) but is adapted for a CP application. 
Proposed Sec.  53.1312(b)(1) is equivalent to Sec.  52.79(b), (c), and 
(d) but is adapted for a CP application. Section 53.1312(b)(2) is 
equivalent to portions of Sec. Sec.  52.63(b)(1), 52.79(b)(1) through 
(b)(3), (c), and (d)(1) and (d)(3), 52.80, and 52.93(b), but is adapted 
for a CP application. Guidance for equivalent requirements in parts 50 
and 52 is also addressed in RG 1.206, ``Applications for Nuclear Power 
Plants,'' Revision 1, section C.1.7.
    Proposed Sec.  53.1315 would address standards for review of 
applications and administrative review of applications, including 
hearings, and is equivalent to Sec. Sec.  52.81 and 52.85, but is 
adapted for a CP application.
    Proposed Sec.  53.1318 would address finality of NRC approvals, 
licenses, and certifications referenced in a CP application and is 
equivalent to Sec.  52.83(a) but is adapted for a CP application.
    Proposed Sec.  53.1324 would address referral to the ACRS and is 
equivalent to Sec.  50.58(a) and to Sec.  52.87 but is adapted for a CP 
application.
    Proposed Sec.  53.1327 would address authorization to conduct LWA 
activities and is equivalent to Sec.  52.91 but is adapted for a CP 
application. Proposed Sec.  53.1327(a) is equivalent to Sec.  52.91(a) 
but is adapted for a CP application. Proposed Sec.  53.1327(b) is 
equivalent to Sec.  52.91(b) but is adapted for a CP application. 
Proposed Sec.  53.1330 would address exemptions, departures, and 
variances for CP applicants.
    Proposed Sec.  53.1333 would address issuance of CPs. Proposed 
Sec.  53.1333(a) is equivalent to Sec.  50.35(a). Proposed Sec.  
53.1333(b) is equivalent to Sec.  50.35(b) and to Sec.  52.97(c) but is 
adapted for a CP application. Proposed Sec.  53.1336 would address the 
effect of CPs and is equivalent to Sec.  50.35(b). Proposed Sec.  
53.1342 would address the duration of CPs. Proposed Sec.  53.1342(a) is 
equivalent to Sec.  50.55(a). Proposed Sec.  53.1342(b) is equivalent 
to Sec.  50.55(b). Proposed Sec.  53.1345 would address the transfer, 
assignment, and disposal of CPs and is equivalent to Sec.  50.80. 
Proposed Sec.  53.1348 would address the termination of CPs and is 
equivalent to Sec. Sec.  52.3(b)(8) and 52.110(a)(1) but is adapted for 
a CP application.
    Proposed Sec. Sec.  53.1360 through 53.1405 address requirements 
for OLs.
    Proposed Sec.  53.1366 would address requirements for the general 
content of applications for OLs. It would refer to general content 
requirements in proposed Sec.  53.1109 and would require supplemental 
information. Proposed Sec.  53.1366(a) is equivalent to Sec.  50.33(f). 
Proposed Sec.  53.1366(b) is equivalent to Sec.  50.33(k).
    Proposed Sec.  53.1369 would provide requirements for the technical 
content of applications for OLs to be included in the FSAR and is 
equivalent to Sec.  50.34(b) but has been modified to align with the 
technical requirements in part 53. It would require that the FSAR 
include and, as needed, update information provided in the PSAR that 
was submitted and reviewed to support the associated CP application.
    Similar to the proposed requirements for the content of CP 
applications, proposed Sec.  53.1369(a) would reference the 
requirements for the content of an ESP application to address 
application requirements related to the site. Section 53.1369(b) would 
reference the requirements for the content of a DC application to 
address some of the application requirements related to design of the 
commercial nuclear plant.
    Proposed Sec.  53.1369(c) is equivalent to Sec.  50.34(b)(7). 
Proposed Sec.  53.1369(d) would require a description of the Integrity 
Assessment Program that would be required by proposed Sec.  53.870. 
Proposed Sec.  53.1369(e) is equivalent to Sec.  50.34(e). Proposed 
Sec.  53.1369(g) would provide requirements for OL application content 
to support proposed Sec.  53.730 related to the role of personnel in 
the operation of the commercial nuclear plant and is adapted from 
requirements in part 55 and Sec.  50.34(f). Likewise, proposed Sec.  
53.1369(h) would provide

[[Page 86947]]

requirements for OL application content related to training programs to 
support proposed Sec. Sec.  53.730(g) and 53.830 and includes 
requirements equivalent to Sec.  50.34(b)(8), Sec.  52.79(a)(33), and 
part 55. Proposed Sec.  53.1369(i) would provide requirements for OL 
application content related to emergency plans to support proposed 
Sec.  53.855 and is equivalent to Sec.  50.34(b)(6)(v).
    Proposed Sec.  53.1369(j) would provide requirements for OL 
application content related to the applicant's organizational structure 
and is equivalent to Sec.  50.34(b)(6)(i). Proposed Sec.  53.1369(k) 
would provide requirements for OL application content related to the 
applicant's proposed maintenance program to support proposed Sec.  
53.715 and is equivalent to Sec.  50.34(b)(6)(iv). Proposed Sec.  
53.1369(l) would provide requirements for OL application content 
related to the applicant's quality assurance program to support 
proposed Sec.  53.865 and is equivalent to Sec.  50.34(b)(6)(ii). 
Proposed Sec.  53.1369(m) would provide requirements for OL application 
content related to the applicant's proposed radiation protection 
program to support proposed Sec.  53.850 and is equivalent to Sec.  
50.34(b)(3).
    Proposed Sec.  53.1369(n) through (p) would provide requirements 
for OL application content related to the applicant's proposed physical 
security program to support proposed Sec.  53.860(a) and are equivalent 
to Sec.  50.34(c) and (d). Proposed Sec.  53.1369(q) would provide 
requirements for OL application content related to the applicant's 
proposed cybersecurity plan to support proposed Sec.  53.860(d) and is 
equivalent to Sec. Sec.  52.79(a)(36)(iv) and 73.54. Proposed Sec.  
53.1369(r) would provide requirements for OL application content 
related to the implementation of proposed security, safeguards, and 
cybersecurity plans to support proposed Sec.  53.860 and is equivalent 
to Sec.  52.79(a)(35)(ii) and 52.79(a)(36)(iv) and (v).
    Proposed Sec.  53.1369(s) would provide requirements for OL 
application content related to the applicant's proposed fire protection 
program to support proposed Sec.  53.875 and is equivalent to Sec.  
52.79(a)(40). Proposed Sec.  53.1369(t) would provide requirements for 
OL application content related to the applicant's proposed ISI and IST 
program to support proposed Sec.  53.880 and is equivalent to part of 
Sec.  52.79(a)(11). Proposed Sec.  53.1369(w) would provide 
requirements for OL application content related to the applicant's 
general employee training program to support proposed Sec.  53.830 and 
is equivalent to Sec.  52.79(a)(33). Proposed Sec.  53.1369(x) would 
provide requirements for OL application content related to the 
applicant's FFD program to support part 26 and is equivalent to Sec.  
52.79(a)(44). Proposed Sec.  53.1369(y) would provide requirements for 
OL applicant's programs to demonstrate that any safety questions 
identified at the CP stage have been resolved and is equivalent to 
Sec.  50.34(b)(5). Proposed Sec.  53.1369(z) would provide requirements 
for OL applicants to describe how the performance of each safety design 
feature has been demonstrated capable of fulfilling functional design 
criteria considering interdependent effects through either analysis, 
appropriate test programs, prototype testing, operating experience, or 
a combination thereof to support proposed Sec.  53.440(a). It is 
largely equivalent to Sec. Sec.  50.34(b)(5) and 50.43(e). Proposed 
Sec.  53.1369(aa) would provide requirements for OL application content 
related to the applicant's proposed TS to support proposed Sec.  
53.710(a) and is equivalent to Sec.  50.34(b)(6)(vi).
    Proposed Sec.  53.1372 would address requirements for the content 
of OL applications other than the FSAR. Proposed Sec.  53.1372(a) would 
require submission of an environmental report and is equivalent to 
Sec.  50.30(f) and Sec.  51.53(b). Proposed Sec.  53.1372(b) does not 
have a direct parallel in parts 50 and 52 and would require the 
inclusion of a description of availability controls that are not 
included in the FSAR to support proposed Sec.  53.710(b).
    Proposed Sec.  53.1375 would address standards for review of OL 
applications and the administrative review of applications, including 
hearings, and is equivalent to Sec. Sec.  52.81 and 52.85, except that 
the NRC has omitted 10 CFR part 54, ``Requirements for Renewal of 
Operating Licenses for Nuclear Power Plants,'' from the list of 
standards in the proposed Sec.  53.1375(a). Proposed part 53 does not 
include detailed requirements related to renewal of licenses, although 
a general provision and possible placeholder for future requirements 
has been included as proposed Sec.  53.1595. The NRC will decide after 
the part 53 final rule is published whether this future section will be 
retained in part 53 to address license renewal or whether the agency 
will take another approach to address license renewal for part 53 
licensees, such as amending part 54 to address part 53 licensees.
    Proposed Sec.  53.1381 would address referral to the ACRS and is 
equivalent to Sec. Sec.  50.58 and 52.87. Proposed Sec.  53.1384 would 
address exemptions, departures, and variances for OL applicants. 
Section 53.1384(a) is equivalent to Sec.  52.93 but is adapted for OLs. 
Proposed Sec.  53.1384(b) is equivalent to Sec. Sec.  52.39(d) (with 
respect to ESPs) and 52.93 but is adapted for OLs.
    Proposed Sec.  53.1387 would address issuance of OLs. The proposed 
introductory paragraph is equivalent to Sec.  50.56. Proposed Sec.  
53.1387(a)(1)(i) is equivalent to Sec. Sec.  50.50 and 50.57(a)(1). 
Proposed Sec.  53.1387(a)(1)(ii) is equivalent to Sec.  50.50. Proposed 
Sec.  53.1387(a)(1)(iii) is equivalent to Sec.  50.57(a)(2). Section 
53.1387(a)(1)(iv) is equivalent to Sec.  50.57(a)(3). Proposed Sec.  
53.1387(a)(1)(v) is equivalent to Sec.  50.57(a)(4). Proposed Sec.  
53.1387(a)(1)(vi) is equivalent to Sec.  50.57(a)(6). Proposed Sec.  
53.1387(a)(1)(vii) is equivalent to Sec.  50.57(a)(5). Proposed Sec.  
53.1387(a)(1)(viii) is equivalent to Sec.  52.97(a)(1)(vi) but is 
adapted for OLs. Proposed Sec.  53.1387(c) is equivalent to Sec.  
50.57(b). Proposed Sec.  53.1387(d) is equivalent to Sec. Sec.  
50.36(b) and 50.50.
    Proposed Sec.  53.1390 would address backfitting of OLs and is 
equivalent to Sec.  52.98(a) but adapted for an OL application. 
Proposed Sec.  53.1396 would address duration of an OL and is 
equivalent to Sec.  50.51(a) and Sec.  52.104. Proposed Sec.  53.1399 
would address transfer, assignment, and other disposition of an OL and 
is equivalent to Sec.  50.80. Proposed Sec.  53.1402 would address 
applications for renewal of an OL and refers to proposed Sec.  53.1595. 
Proposed Sec.  53.1405 would address continuation of an OL and is 
equivalent to Sec.  52.109 but is adapted to address an OL.
    Proposed Sec. Sec.  53.1410 through 53.1461 would address 
requirements for COLs. Proposed Sec.  53.1410 is equivalent to Sec.  
52.71. Proposed Sec.  53.1413 would address general information 
requirements for the content of applications for COLs and is equivalent 
to Sec.  52.77, which references Sec.  50.33. Most of the provisions 
from Sec.  50.33 are restated in proposed Sec.  53.1109. Some 
requirements in Sec.  50.33 related to financial qualifications and 
construction timelines are addressed in other sections of part 53.
    Proposed Sec.  53.1416 would address the technical content to be 
included in an FSAR for an application for a COL and is equivalent to 
Sec.  52.79 except as modified to reflect the technical requirements in 
part 53 and with one addition. Proposed Sec.  53.1416 includes the 
statement that the Commission will require, before issuance of a COL, 
that engineering documents, such as analyses, drawings, procurement 
specifications, or construction and installation specifications, be 
completed and available for audit if the more

[[Page 86948]]

detailed information is necessary for the Commission to verify the 
information in the application and make its safety determination. This 
statement is equivalent to DC application requirements in Sec.  52.47 
and is included in proposed Sec.  53.1416 for clarity.
    Similar to the proposed requirements for the content of OL 
applications, proposed Sec.  53.1416(a)(1) would reference the 
requirements for the content of an ESP application to address 
application requirements related to siting. Section 53.1416(a)(2) would 
reference the requirements for the content of a DC application to 
address some of the application requirements related to design of the 
commercial nuclear plant. The remaining items under proposed Sec.  
53.1416(a) are likewise similar to the required content for OL 
applications under proposed Sec.  53.1369(a). Proposed Sec.  53.1416(b) 
would require COL applicants to provide a report documenting the 
resolution of any safety questions for SSCs for which research and 
development was necessary to confirm the adequacy of their design and 
is equivalent to Sec.  50.34(b)(5). Proposed Sec.  53.1416(c) would 
provide requirements for COL applicants to describe how the performance 
of each safety design feature has been demonstrated capable of 
fulfilling functional design criteria considering interdependent 
effects through either analysis, appropriate test programs, prototype 
testing, operating experience, or a combination thereof to support 
proposed Sec.  53.440(a). It is largely equivalent to Sec. Sec.  
52.79(a)(24) and 50.43(e). Proposed Sec.  53.1416(d) would address the 
content of COL applications referencing an ESP. Proposed Sec.  
53.1416(e) would address the content of COL applications referencing a 
standard design approval. Proposed Sec.  53.1416(f) would address the 
content of COL applications referencing a standard DC. Proposed Sec.  
53.1416(g) would address the content of COL applications referencing an 
ML.
    Proposed Sec.  53.1419 would address other application content for 
COLs and is equivalent to Sec.  52.80. Proposed Sec.  53.1419(a)(2) is 
new and would require the inclusion of a description of availability 
controls that are not required to be included in the FSAR.
    Proposed Sec.  53.1422 would address standards for review of 
applications and the administrative review of applications, including 
hearings, and is equivalent to Sec. Sec.  52.81 and 52.85. The NRC has 
removed part 54 from the list of standards in proposed Sec.  
53.1422(a). Proposed part 53 does not include requirements related to 
renewal of licenses, in relation to proposed Sec. Sec.  53.1422 and 
53.1595.
    Proposed Sec.  53.1425 would address the finality of NRC approvals 
referenced in a COL application and is equivalent to Sec.  52.83(a). 
Proposed Sec.  53.1431 would address the referral of COL applications 
to the ACRS for review and is equivalent to Sec.  52.87. Proposed Sec.  
53.1434 would address the authorization to conduct LWA activities and 
is equivalent to Sec.  52.91. Proposed Sec.  53.1437 would address 
exemptions, departures, and variances and is equivalent to Sec.  52.93. 
Proposed Sec.  53.1440 would address issuance of COLs and is equivalent 
to Sec.  52.97. Proposed Sec.  53.1443 would address finality of COLs 
and is equivalent to Sec.  52.98.
    Proposed Sec.  53.1449 would address inspection during construction 
and is equivalent to Sec.  52.99. Proposed Sec.  53.1452 would address 
operation under a COL and is equivalent to Sec.  52.103. Paragraph (a) 
of proposed Sec.  53.1452 would include footnotes to provide that, for 
licensees installing fueled manufactured reactors under a COL, (1) the 
COL holder would notify the NRC of its scheduled date for initiating 
the physical removal of any one of the independent physical mechanisms 
to prevent criticality required under Sec.  53.620(d)(1) rather than 
its scheduled date for the initial loading of fuel, and (2) the NRC 
would time its publication of the notice of intended operation based on 
the COL holder's schedule for initiating the physical removal of any 
one of the independent physical mechanisms to prevent criticality 
required under Sec.  53.620(d)(1) rather than the COL holder's 
scheduled date for the initial loading of fuel. These footnotes are 
consistent with the provisions of proposed Sec.  53.620(d)(1)(iv), 
which would state that, upon initiating the physical removal of any one 
of the independent physical mechanisms to prevent criticality in the 
manufactured reactor's place of operation, the fueled manufactured 
reactor has commenced operation. For reactors without the independent 
physical mechanisms to preclude criticality under proposed Sec.  
53.620(d)(1), operation begins with initial fuel load. In both cases, 
removal of the physical features to prevent criticality (for reactors 
with such features) and initial fuel load (for reactors without such 
features) put a fully constructed utilization facility in a position to 
sustain a nuclear chain reaction, and in both cases, the utilization 
facility cannot sustain a nuclear chain reaction (for lack of 
sufficient reactivity) until the action takes place. Therefore, the NRC 
proposes that initiating the physical removal of any one of the 
independent physical mechanisms to prevent criticality is the best 
analogue to initial loading of fuel for reactors without such features.
    The proposed footnote in Sec.  53.1452(a) regarding timing of the 
notice of intended operation for fueled manufactured reactors with 
independent physical mechanisms to prevent criticality also addresses 
the requirements of Section 189a.(1)(B)(i) of the Act. This section 
requires, in part, that ``[n]ot less than 180 days before the date 
scheduled for initial loading of fuel into a plant by a licensee that 
has been issued a combined construction permit and operating license 
under section 185b., the Commission shall publish in the Federal 
Register notice of intended operation.'' That section further requires 
that this notice provide a 60-day period in which to request a hearing 
``on whether the facility as constructed complies, or on completion 
will comply, with the acceptance criteria of the license.'' In the case 
where a fueled manufactured reactor arrives at the site where it is to 
be operated by a COL holder, the manufacturer would have loaded fuel at 
the factory under its part 70 license. Therefore, at the site of 
operation, there would not be ``initial loading of fuel into a plant by 
a licensee that has been issued a combined construction permit and 
operating license'' (emphasis added). Under a literal reading of the 
entry condition in Act Section 189a.(1)(B)(i), this situation would not 
trigger its requirements. However, the purpose of the provision is to 
offer the hearing opportunity at least 180 days prior to when the fuel 
is loaded and ready for use at its authorized location. It would be 
contrary to that purpose if, in this situation, the Commission did not 
publish the notice of intended operation and opportunity for the public 
to request a hearing on conformance with the acceptance criteria in the 
COL for the site of operation. To fulfill the underlying purpose of the 
law, the NRC proposes to time the notice of intended operation based on 
the COL holder's schedule for initiating the physical removal of any 
one of the independent physical mechanisms to prevent criticality 
required under Sec.  53.620(d)(1). This action by the COL holder would 
be the best analogue to initial fuel load by the COL holder for the 
reasons stated previously. This analogue is adopted in other sections 
of the proposed part 53 and related sections in parts 50 and 73 that 
use initial fuel loading to identify

[[Page 86949]]

a transition point for the applicability of regulatory requirements. To 
address the possible loading of fuel into a manufactured reactor for 
subsequent transport to and use at a commercial nuclear plant, multiple 
sections that determine the applicability of regulations have been 
drafted or revised to allow for either initial fuel load or initiating 
the physical removal of any one of the independent physical mechanisms 
to prevent criticality required under Sec.  53.620(d)(1) for a fueled 
manufactured reactor to determine the applicability of the requirement, 
as appropriate.
    Proposed Sec.  53.1455 would address duration of COL and is 
equivalent to Sec.  52.104. Proposed Sec.  53.1456 would address the 
transfer of a COL and is equivalent to Sec.  52.105. Proposed Sec.  
53.1458 would address application for renewal and is equivalent to 
Sec.  52.107. Proposed Sec.  53.1461 would address continuation of COL 
and is equivalent to Sec.  52.109.
    Proposed Sec.  53.1470 would address standardization of commercial 
nuclear plant designs and licenses to construct and operate commercial 
power reactors of identical design at multiple sites and is equivalent 
to appendix N of part 52. This section would set out the particular 
requirements and provisions applicable to situations in which 
applications for CPs and subsequent OLs, or COLs, under this part are 
filed by one or more applicants for licenses to construct and operate 
nuclear power reactors of identical design (``common design'') to be 
located at multiple sites. Additional information related to this 
proposed section is provided in the final rule to revise part 52 (72 FR 
49352; August 28, 2007).

Subpart I--Maintaining and Revising Licensing-Basis Information

    Part 53 would establish requirements for the maintenance of 
licensing-basis information in subpart I.
    Section 53.1500 would describe the purpose of the subpart in terms 
of the definition of licensing-basis information in subpart A. Subpart 
I would be closely tied to the requirements in subpart H, which would 
provide the requirements for contents of applications for the various 
types of licenses issued under part 53. Subpart I would generally be 
organized into sections dealing with: (1) licensing-basis information 
that licensees are not authorized to change without NRC approval (e.g., 
licenses, regulations); and (2) licensing-basis documents that 
licensees may change provided specified criteria are satisfied (e.g., 
FSAR, program descriptions). The subpart would also capture certain 
general conditions on licenses and changes to the licenses related to 
the transfer and termination of licenses.
    Section 53.1502 would define specific terms and conditions of 
licenses. These terms and conditions would be equivalent to the 
regulations in: (1) Sec.  50.54(h) stating that each license is subject 
to the provisions of the Act and requirements issued by the Commission; 
(2) Sec.  50.54(s) stating the actions the Commission would take if it 
makes a finding that there is not reasonable assurance that adequate 
protective measures can and will be taken in the event of a 
radiological emergency; (3) Sec.  50.54(aa) stating that each license 
is subject to the specified sections of the Federal Water Pollution 
Control Act; and (4) Sec.  50.54(dd) stating that a holder of an OL or 
COL may take reasonable actions that depart from the license in a 
national security emergency.
    Section 53.1505(a) would serve as an introduction to and overview 
of the sections that follow on changes to licensing-basis information 
requiring prior NRC approval, namely the elements of licensing-basis 
information defined by licenses, orders, and regulations. The related 
sections within these subparts would primarily deal with the process of 
how a licensee requests and the NRC issues an amendment to a license or 
issues an order that modifies a license. Another important element of 
licensing-basis information that a part 53 licensee would not be able 
to change or deviate from without NRC approval would be the NRC 
regulations themselves. Section 53.1505(b) would refer to Sec.  53.080 
in subpart A that would provide the criteria for a licensee or other 
party to satisfy when requesting an exemption from NRC regulations.
    Section 53.1510 would be equivalent to Sec.  50.90 and would 
require that a licensee submit an application to request an amendment 
to a license. The required assessments that would be included within an 
application to amend a license under part 53 would need to address the 
safety criteria and analysis requirements of subparts B and C. As with 
parts 50 and 52, licensees would be required to include in their 
applications to amend a license an analysis of whether the amendment 
involves no significant hazards consideration using the standards in 
Sec.  53.1520, which would be equivalent to the standards in Sec.  
50.92. Although this rulemaking provided an opportunity to revise the 
terminology related to no significant hazards consideration 
determinations, which dates to the early 1960s when applications were 
supported by final hazard summary reports, the NRC is proposing to 
maintain the same terminology used in part 50 to minimize the need for 
associated changes in other regulations, guidance, and public notices.
    Section 53.1515 would establish requirements for public notices and 
state consultations associated with the NRC's processing of a license 
amendment request. This section would be equivalent to Sec.  50.91 for 
the NRC's processes related to applications to amend an OL or COL. 
Section 50.91(b) stipulates that the Commission will make available to 
the licensee the name of the appropriate State official designated to 
receive such amendments. While the Commission intends to continue 
following this practice, the Commission has not included this 
administrative matter in proposed part 53. Proposed Sec.  53.1515(b)(3) 
contains some modifications compared to Sec.  50.91(b)(3) for clarity; 
these revisions are not intended to revise the substance of the 
provisions in part 53 compared to part 50.
    Section 53.1520 would be based on Sec.  50.92. The section would 
continue to use the criteria in Sec.  50.92 for determining that a 
proposed amendment involves no significant hazards consideration. 
Although more specific terms such as event sequence are used throughout 
part 53, Sec.  53.1520 would use the term ``accident'' to maintain 
consistency with the long history of making no significant hazards 
consideration determinations under part 50.
    Section 53.1525 would provide requirements for holders of an OL or 
COL requesting to revise information from a DC rule that was referenced 
in the initial license application and included in or incorporated by 
reference into the facility FSAR. In keeping with the current 
requirements in part 52, the portion of the part 53 facility licensing-
basis information obtained from the certified design would be divided 
into two categories. The most significant design information and the 
ITAAC would be certified by rule and designated as ``certification 
information.'' The remaining information, which makes up the majority 
of the design information approved as part of the DC, would not be 
certified by rule and is not considered ``certification information.'' 
Part 52 refers to these categories of information as Tier 1 and Tier 2 
information, respectively, and refers to a change made to that 
information on a plant-specific basis as a departure. Under part 52, a 
departure from Tier 1 information requires an exemption and,

[[Page 86950]]

for information incorporated into the license, a license amendment.
    Part 53 would dispense with the Tier 1 and Tier 2 terminology. 
Rather, Sec.  53.1525 would use the term ``certification information'' 
in place of Tier 1, and a plant-specific departure from the 
certification information would require both a request for an exemption 
from the associated DC rule and, for information such as ITAAC 
incorporated into the license, a license amendment. However, as would 
be provided in Sec.  53.1525(c), a plant-specific departure from the 
information approved by the NRC as part of the DC rule but which is not 
certification information (i.e., Tier 2 information under part 52) 
would be assessed using the process and criteria defined in Sec.  
53.1550 for changes to a FSAR. An applicant or licensee would need to 
identify such a change as a departure from the referenced standard 
design in the updated FSAR. The process for making a generic change to 
a certified design would be described in the associated section in 
subpart H.
    Section 53.1530 would not allow the holder of an ML or the holder 
of a COL using a manufactured reactor to make changes to the design of 
the manufactured reactor without requesting a license amendment from 
the NRC. This section would provide the equivalent requirements as 
those in Sec. Sec.  52.98 and 52.171.
    Section 53.1535 would establish requirements for license amendments 
during construction. The section would provide the equivalent options 
and requirements for the holders of a CP as those in Sec.  50.35(b). 
The regulations would allow but do not require the holder of a CP or 
LWA to request an amendment under Sec.  53.1510 if the licensee desires 
to obtain NRC approval of a specific design feature or specification. 
The requirements for obtaining an amendment to a COL to address changes 
during construction would also be provided in Sec.  53.1535. The 
proposed process would differ from the current requirements in part 52 
by adopting a requirement that would explicitly support a change 
process like that described in RG 1.237, ``Guidance for Changes During 
Construction for New Nuclear Power Plants Being Constructed Under a 
Combined License Referencing a Certified Design Under 10 CFR part 52.''
    The proposed regulation would allow the holder of a COL to proceed 
at its own risk in making a change during the construction process and 
would require that licensee to submit a license amendment request no 
later than 45 days from the date the licensee begins to implement the 
change or departure requiring NRC approval.
    Section 53.1540 would serve as an introduction to the sections that 
follow on changes to licensing-basis information that are primarily 
under the control of a licensee but for which evaluations are made to 
determine if a submittal to the NRC requesting approval would be 
required. The section would also include definitions that would be 
applicable when using the processes in Sec. Sec.  53.1545 through 
53.1565. The definitions would be largely equivalent to those in Sec.  
50.59(a) but include some revision to reflect the structure and 
terminology in other subparts in part 53. For example, the definition 
of ``Change'' in Sec.  53.1540(b) would address a ``design feature or 
related functional design criteria'' rather than a ``design function,'' 
because the former are defined terms in part 53. Similarly, in Sec.  
53.1540(b), the phrase ``design basis'' from Sec.  50.59(a)(2) would be 
replaced with functional design criteria for SR SSCs.
    Section 53.1545 would provide the proposed requirements for 
updating of FSARs. While the process-related requirements proposed 
under Sec.  53.1545 would be largely the same as those in Sec.  50.71, 
the specifics of information to be updated would differ due to the role 
of PRA in satisfying the requirements in subparts B and C. 
Additionally, the use of the risk-informed approach in subpart C would 
result in some but not all PRA information being in the FSAR or another 
licensing basis document and therefore a separate PRA update 
requirement similar to Sec.  50.71(h) is not included in proposed 
subpart I.
    Proposed Sec.  53.1239(a)(18) in subpart H and the related 
references to this proposed requirement for the holders of OLs and COLs 
would require a description of the PRA required by Sec.  53.450(a) and 
its results to be included in FSARs. However, guidance documents are 
planned to clarify the division of PRA-related information that would 
need to be in the FSAR, in other possible licensing basis documents, 
and controlled as plant records subject to inspections and audits. At a 
minimum, the information from the PRA that would be needed to show 
compliance with subpart C would be included in the FSAR (e.g., PRA 
summary and analytical results for LBEs). The submittal of voluminous 
PRA information was initially required under part 52, but that proved 
to be impractical and was revised in the 2007 revision of part 52. 
Guidance is being developed to ensure sufficient information is 
submitted to the NRC to support the licensing process and the NRC's 
regulatory findings under part 53 or similar applications using the LMP 
under parts 50 or 52.
    The NRC has posed a question in section VI, ``Specific Requests for 
Comments,'' of this document that asks about the appropriate level of 
detail for PRA-related information in an FSAR and whether other 
licensing basis documents might be more appropriate to both provide 
information to the NRC and ensure the PRA is maintained and updated as 
proposed in subpart C. The program document would provide more detail 
than the summaries in the FSAR but still be a much-condensed source of 
information in comparison to the documentation of the PRA.
    Section 53.1545(a)(3) and (4) would be based on the inclusion of at 
least a summary of PRA results and the related margins to safety 
criteria in the FSAR and would require updates to that information. The 
routine reporting of these margins would also inform application of the 
criteria for allowing changes without an amendment in the following 
section (Sec.  53.1550) in subpart I.
    Section 53.1550 would establish requirements for evaluating changes 
to a facility as described in its FSAR. This proposed section would 
provide the equivalent of the requirements in Sec.  50.59 for 
evaluating changes to an FSAR (as updated) and determining if a license 
amendment is required to implement a change to a facility or 
procedures. The evaluation criteria proposed in Sec.  53.1550 would 
reflect the role of the PRA in the safety analyses under part 53 and 
would include several measures related to the changes in plant risk 
resulting from a change in the plant design or plant procedures. 
Examples would include criteria that rely on the identification of 
risk-significant event sequences in accordance with the analysis 
requirements of Sec.  53.450; exceeding the LBE evaluation criteria as 
defined in Sec.  53.450; the consideration of potential reductions in 
margin between the estimated comprehensive risk metrics and associated 
risk performance objectives in the safety criteria in Sec.  53.220; 
changes to the safety classification of SSCs; and consideration of 
reductions in defense in depth.
    Section 53.1550 would include a criterion related to a departure 
from a method of evaluation used in the safety analyses. The NRC has 
not yet developed draft guidance for use in applying proposed Sec.  
53.1550 but anticipates that the NRC and stakeholders will assess the 
potential need for such guidance and that such guidance would, if 
needed, be

[[Page 86951]]

developed as part of ongoing or future activities.
    Section 53.1550 would include certain concepts taken from existing 
guidance for Sec.  50.59 in the proposed criteria related to DBAs. 
Specifically, criterion (iv) for changes made to a method of evaluation 
of DBAs under Sec.  53.450(f) would be equivalent to a change in a 
method of evaluation under Sec.  50.59, and criterion (viii) on 
assessing if a change creates a possibility for an accident of a 
different type than previously analyzed in the FSAR would be similar to 
the Sec.  50.59 criterion (v). Guidance documents will be prepared to 
address the content of applications for PRA-related information under 
proposed part 53, and this guidance will also influence how potential 
changes in the evaluation of LBEs other than DBAs analyzed under Sec.  
53.450(e) are evaluated and reported under the proposed criterion (iv).
    Section 53.1550(a)(2)(x) would require evaluating plant changes to 
ensure they would not prevent satisfying the design requirements in 
Sec.  53.440(j) related to the impact of a large commercial aircraft. 
The inclusion of a proposed requirement under Sec.  53.1550 related to 
design features for protecting against aircraft impact would reflect 
the proposed design requirement in subpart C and related proposed 
requirements in subpart H to address the proposed design requirement in 
FSARs.
    Sections 53.1560 through 53.1565 in subpart I would define the 
processes for a licensee to evaluate changes to the program documents 
included in the licensing-basis information submitted to the NRC and to 
modify such programs without NRC prior approval.
    Section 53.1560 would include the proposed requirements for 
updating program documents included in licensing-basis information and 
would provide the equivalent of FSAR updates for key program documents. 
The proposed requirements in these sections would provide a uniform 
approach for updating program documents, which correspond to the 
programs required under subpart F.
    The proposed Sec.  53.1565 would provide a process for licensees to 
make changes to program documents included in licensing-basis 
information without obtaining prior NRC approval. The proposed 
requirements would include several generic criteria that, if not 
satisfied, would prompt the need for NRC approval of a change to a 
program document. These generic criteria would include whether a change 
would comply with TS and NRC regulations. Another proposed criterion 
for evaluating changes to program documents would be conforming with 
program-specific requirements, including NRC-approved program documents 
with more specific criteria for a particular program, regulations, 
administrative controls sections of TS, and NRC-approved program 
documents.
    Proposed Sec.  53.1565(d) would include specific criteria for 
evaluating changes to several program documents that have well 
established change processes and guidance for licensees under parts 50 
and 52. The program documents specifically addressed in the proposed 
section would include quality assurance programs that would be 
equivalent to Sec.  50.54(a), an emergency preparedness program that 
would be equivalent to Sec.  50.54(q), and the security program that 
would be equivalent to Sec.  50.54(p).
    The proposed Sec.  53.1570 would establish requirements for the 
transfer of commercial nuclear plant licenses by providing the 
equivalent requirements of Sec.  50.80 for the possible transfer of an 
ESP, CP, OL, or COL. Likewise, the proposed Sec.  53.1575 would 
establish requirements for the termination of an OL or COL by providing 
the equivalent requirements of Sec.  50.82. Other proposed requirements 
related to decommissioning and license termination would be included in 
subpart G.
    Section 53.1580 would establish requirements for information 
requests the NRC could send to the various types of licensees and would 
provide requirements that would be equivalent to requirements in Sec.  
50.54(f). The proposed Sec.  53.1585 would provide the requirements 
that would be equivalent to requirements in Sec.  50.100 to address 
revocation, suspension, modification of licenses, and approvals for 
cause. Section 53.1590 would propose to address backfitting 
requirements by providing requirements that would be equivalent to 
those in Sec.  50.109.
    Proposed Sec.  53.1595 would address license renewals under part 53 
with simple statements that licenses may be renewed. This section would 
be expanded through future rulemakings to more fully describe or 
reference the processes related to requesting and processing 
applications to renew ESPs, OLs, and COLs issued under part 53 (if 
finalized).

Subpart J--Reporting and Other Administrative Requirements

    Part 53 would address various reporting and administrative 
requirements in subpart J.
    Section 53.1600 would explain the organization of the various 
sections within the subpart related to providing unfettered access to 
NRC inspectors; maintaining certain records and reporting specified 
events or conditions; demonstrating compliance with financial 
qualification requirements and providing specified financial reports; 
and maintaining financial protections to address potential accidents.
    Section 53.1610 would establish requirements for the provision of 
facilities and unfettered access for inspections. These requirements 
would be equivalent to Sec.  50.70 with only minor changes proposed to 
provide additional flexibilities and address possible differences 
related to reactors licensed under part 53 and the possibility that 
some commercial nuclear plants may not be assigned resident inspectors.
    Section 53.1620 would provide for maintenance of records and the 
making of various reports to the NRC. These requirements would be 
largely equivalent to Sec.  50.71. This section is not intended to 
reflect all provisions in Sec.  50.71; several important requirements 
in Sec.  50.71 would be captured in other sections of part 53. For 
example, Sec.  53.1545 within subpart I would provide requirements that 
would be equivalent to Sec.  50.71(e), updating FSARs, and Sec.  
53.1680, ``Annual financial reports,'' would provide the equivalent of 
Sec.  50.71(b), which covers financial reports. A reporting requirement 
related to completion of power ascension testing would be added to 
Sec.  53.1620 to support the assessment of annual fees under 10 CFR 
part 171, ``Annual Fees for Reactor Licenses and Materials Licenses, 
Including Holders of Certificates of Compliance, Registrations, and 
Quality Assurance Program Approvals and Government Agencies Licensed by 
the NRC,'' which normally commence upon completion of those testing 
activities.
    Section 53.1630 would establish requirements for immediate 
notification requirements for operating commercial nuclear plants. 
These requirements would be equivalent to Sec.  50.72 with minor 
changes proposed to make the reporting criteria technology inclusive. 
In addition, a new version of NRC Form 361 (NRC Form 361S) would be 
created for use by part 53 licensees, but without LWR-specific 
terminology to ensure technology inclusiveness. A separate rulemaking 
activity, ``Reporting Requirements for Nonemergency Events at Nuclear 
Power Plants,'' has been initiated to consider possible changes to the 
requirements in Sec.  50.72. At a future date, the NRC may consider 
reconciling future changes to Sec.  50.72 with the requirements 
proposed in part 53, which have been taken or derived from the current 
reporting requirements.

[[Page 86952]]

    Section 53.1640 would address the licensee event report system. 
These requirements would be equivalent to Sec.  50.73 with minor 
changes proposed to make the requirements inclusive of various reactor 
technologies and to reflect appropriate internal references to other 
sections in part 53. In addition, NRC Forms 366, 366A, and 366B would 
be revised to include corresponding check boxes for part 53 licensees.
    Section 53.1645 would require periodic reporting of the quantity of 
radionuclides released to unrestricted areas in liquid and gaseous 
effluents, doses to members of the public, and the results of 
environmental monitoring. These reporting requirements in the proposed 
part 53 would be largely equivalent to those in the TSs required by 
Sec.  50.36a, ``Technical specifications on effluents from nuclear 
power reactors.'' The only difference would be that a Sec.  50.36a 
requirement to specifically address conditions where the dose to the 
maximally exposed individual could be significantly above design 
objectives would refer to a design objective of 10 mrem/year total 
effective dose equivalent, instead of referring to the design 
objectives in appendix I to part 50. The proposed section would also 
include an equivalent to the reporting requirement in section IV of 
appendix I to part 50 if the radiation exposure to a member of the 
public in any calendar quarter exceeds one-half of the annual ALARA 
design objective.
    Section 53.1650 would include a reporting requirement to support 
safeguards agreements between the United States and the International 
Atomic Energy Agency (IAEA) and would be equivalent to Sec.  50.78.
    Section 53.1660 through 53.1700 would address financial 
requirements and would be largely similar to existing regulations in 
parts 50 and 52. Section 53.1670 would be entitled ``Financial 
qualifications'' and would require applicants other than electric 
utilities to possess or have reasonable assurance of obtaining funds 
for the activities for which the license is being sought. The NRC is 
seeking feedback on these sections and their ramifications for merchant 
plants \3\ in section VI, ``Specific Requests for Comments,'' of this 
document. The remaining financial reports in part 53 would be 
equivalent to Sec.  50.71(b) for annual financial reports, Sec.  50.76 
for a change of status, Sec.  50.54(cc) for the filing of a petition 
for bankruptcy, and Sec.  50.81 for creditor regulations.
---------------------------------------------------------------------------

    \3\ A ``merchant plant'' is a plant licensed to a non-rate-
regulated entity (e.g., a nonutility) that engages in the business 
of production, manufacturing, generating, buying, aggregating, 
marketing, or brokering electricity for sale at wholesale or for 
retail sale to the public.
---------------------------------------------------------------------------

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

Subpart M--Enforcement

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

V. Changes to Other Parts of 10 CFR Chapter I

10 CFR Part 26

A. Introduction

    The NRC is proposing a technology-inclusive, risk-informed, and 
performance-based approach for the application of drug and alcohol 
testing and fatigue management requirements for facilities licensed 
under part 53. The proposed requirements applicable to these 
applicants, licensees, and other entities would be commensurate with 
the radiological consequences presented by the applicants' facilities 
and the operation of these facilities.\4\ The proposed FFD framework 
would consist of a two-tiered graded approach similar to that currently 
in part 26 and an optional third tier for part 53 commercial nuclear 
plants that perform an analysis that demonstrates the facility and its 
operation would satisfy the criterion in proposed Sec.  26.603(c), 
which refers to Sec.  53.860(a). This proposed FFD framework would be 
established in subpart M, ``Fitness for Duty Programs for Facilities 
Licensed Under Part 53,'' of part 26.
---------------------------------------------------------------------------

    \4\ The NRC uses the term ``operation'' in its part 26 
discussion to focus on human performance, namely the necessity of 
individuals to operate, maintain, surveil, and protect the facility 
and respond to operational transients and unlikely event sequences.
---------------------------------------------------------------------------

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

[[Page 86953]]

and other entities identify individuals as users of impairing 
substances and demonstrate compliance with Sec.  26.23, ``Performance 
objectives.''
    The proposed second tier would include all the proposed first-tier 
requirements, plus the more comprehensive set of FFD program 
requirements in current subparts C, ``Granting and Maintaining 
Authorization,'' D, ``Management Actions and Sanctions to be Imposed,'' 
H, ``Determining Fitness-for-Duty Policy Violations and Determining 
Fitness,'' and N, ``Recordkeeping and Reporting Requirements,'' of part 
26. These requirements would be provided in proposed Sec.  26.605(b) 
and would be applicable to licensees and other entities satisfying the 
Sec.  26.603(c) criterion, at their discretion. These requirements 
would also apply to licensees or other entities not satisfying the 
Sec.  26.603(c) criterion that implement an FFD program under subpart M 
of part 26, before the loading of fuel onsite into a reactor vessel; 
before receiving a manufactured reactor; or before operating, testing, 
performing maintenance of, or directing the maintenance or surveillance 
of security-related equipment or equipment that a risk-informed 
evaluation process has shown to be significant to public health and 
safety.
    The second-tier requirements are based on the additional risk 
presented by nuclear reactor assembly, testing, fueling, and operation 
and the necessity for human actions in certain event sequences. The 
inclusion of the current part 26 requirements would align proposed part 
53 FFD and AA program requirements with the current FFD and AA programs 
required for facilities licensed under parts 50 and 52. This approach 
would ensure effective and consistent AA and FFD program implementation 
across the commercial nuclear power industry, thereby ensuring uniform 
requirements for individuals who may perform roles and responsibilities 
for multiple facilities regardless of facility licensure.
    Proposed Sec.  26.604 would offer an alternate option for an 
applicant implementing an FFD program under subpart M of part 26. If 
the applicant demonstrates that the criterion in proposed Sec.  
26.603(c) is met, then the applicant (and the subsequent licensee or 
other entity) must still implement an FFD program described in subpart 
M of part 26; however, drug and alcohol testing would not be required 
unless FFD performance declines or the applicant, licensee, or other 
entity elects to implement drug and alcohol testing. The proposed Sec.  
26.604 requirements are equivalent to those proposed in Sec.  26.605(a) 
except for required drug and alcohol testing. This proposed framework 
would focus on the human performance of individuals while they are 
performing those duties and responsibilities that make them subject to 
the FFD program. This performance would be verified through behavioral 
observation, evaluation of any FFD concerns, performance monitoring, 
fatigue management, and determinations of fitness. Applicants that do 
not satisfy the criterion in proposed Sec.  26.603(c), or elect not to 
perform the analysis required to demonstrate that the criterion in 
Sec.  26.603(c) is met, would be subject to an FFD program described in 
Sec.  26.605, ``FFD program requirements for facilities that do not 
implement Sec.  26.604,'' or an FFD program that implements all part 26 
requirements, except for those requirements in subparts K and M of part 
26.
    In establishing the minimum FFD program requirements in Sec.  
26.604, the NRC reviewed current advanced reactor designs against that 
of a non-power production or utilization facility (NPUF) that is not 
required to implement an FFD program for those individuals who have 
unescorted access to the controlled access area (and vital area for 
some facilities), including NRC-licensed operators.\5\ This review was 
performed because commercial nuclear plants licensed under part 53 
could be designed with similar power levels and radiological 
consequences as the currently licensed NPUFs. From this review, three 
principal considerations supported the minimum set of requirements for 
the Sec.  26.604 FFD program.
---------------------------------------------------------------------------

    \5\ Controlled access area and vital area are defined in Sec.  
73.2, ``Definitions.''
---------------------------------------------------------------------------

    First, the radiological consequences presented by a part 53 
licensed facility and its operation that satisfy the criterion in Sec.  
26.603(c) may present a greater potential radiological consequence to 
workers and the public in the vicinity of the facility than does an 
NPUF. Second, the operating characteristics of a part 53 licensed 
facility are unlike that of an NPUF because there may be a higher 
reliance on individuals at the part 53 site to safely and competently 
operate, maintain, surveil, and secure SSCs that may not be required at 
an NPUF, such as systems that provide secondary heat transfer, reactor 
coolant flow, pressure control, and at-power core refueling. 
Differences in operating characteristics could include, for example: 
long-term, full power operation with automated reactivity control 
systems for load-following; active and passive safety and security 
systems; innovative non-light-water heat transfer systems; and energy 
storage and hazardous chemical systems. The individuals at part 53 
facilities may also be required to communicate to individuals both 
onsite and offsite, such as electrical load dispatchers, any conditions 
adverse to safety, security, or quality. Third, part 53 licensed 
facilities may be sited in geographically remote locations that may not 
have a physically available administrative or corporate support team to 
provide face-to-face oversight, engineering expertise, and maintenance 
support like that at NPUFs. This places a higher reliance on those 
individuals required at a part 53 facility being fit for duty and 
trustworthy and reliable because a replacement individual may not be 
readily available.
    The NRC proposes to exclude drug and alcohol testing from the 
proposed Sec.  26.604 framework for five reasons: (1) the Sec.  26.23 
performance objectives can be met through effective implementation of 
the defense-in-depth regulatory framework established by behavioral 
observation, reporting of legal actions, the proposed performance 
monitoring and review program (PMRP), FFD training, and requirements 
from the physical protection, AA, cyber protection, and licensed 
operator programs; (2) the PMRP would require the licensee or other 
entity to monitor its FFD program performance (both qualitatively and 
quantitatively) against its historical site performance, fleet-level 
performance, if applicable, and industry performance. The licensee or 
other entity would be required to implement corrective actions if site 
FFD performance meets a licensee- or other entity-established threshold 
or to resolve a finding resulting from a qualitative review or audit in 
a manner that restores performance and corrects root causes, 
contributing causes, or both; (3) the requirements in proposed Sec.  
26.609, ``Behavioral observation,'' are more robust than those in Sec.  
26.407, ``Behavioral observation,'' of subpart K of part 26 and are 
proposed to synchronize with and reinforce the AA behavioral 
observation requirements in Sec.  73.56, ``Personnel access 
authorization requirements for nuclear power plants,'' or the proposed 
requirements under Sec.  73.120, ``Access authorization program for 
commercial nuclear plants''; (4) a part 53 commercial nuclear plant 
that satisfies the Sec.  26.603(c) criterion will be designed, 
operated, and secured with a radiological risk profile that is lower 
than that described in Sec.  53.860(a)(2) and

[[Page 86954]]

perhaps will approach the radiological risk profile of an NPUF (which 
does not implement an FFD program); and (5) the NRC is aware that a 
part 53 commercial nuclear plant could be designed and constructed in 
such a manner to reduce reliance on an onsite security force to protect 
SSCs, NRC-licensed materials, and sensitive information, with enhanced 
capabilities for the detection, assessment, and delay of a DBT 
adversary.
    Regarding fatigue management requirements, work hour controls would 
be required for personnel at utilization and manufacturing facilities 
in accordance with the existing scoping criteria in Sec.  26.4, ``FFD 
program applicability to categories of individuals,'' as revised in 
this proposed rule. The amended Sec.  26.4 also would be used to 
determine whether an individual would be subject to drug and alcohol 
testing. The applicability of these scoping criteria for certain 
individuals (such as operators and maintenance personnel) would be 
determined by the licensee or other entity through its risk-informed 
evaluation process performed to assess the risk significance of the SSC 
upon which work is being performed or directed by the individual. These 
requirements also would be scaled based on the potential radiological 
consequences presented by the facility. However, fatigue management 
would be applied to all individuals subject to the FFD program, similar 
to FFD program implementation by the current fleet of commercial 
nuclear plants because fatigue management is a proactive requirement 
designed to help prevent on-shift impairment through work hour 
scheduling and time off. The behavioral observation program (BOP) would 
be the principal requirement to provide reasonable assurance that 
individuals on shift are not mentally or physically impaired due to 
fatigue, which in any way could adversely affect their ability to 
safely and competently perform their duties.
    The NRC is proposing subpart M of part 26 for facilities licensed 
under part 53, in lieu of subjecting all part 53 licensees to the same 
part 26 requirements that apply to facilities licensed under part 50 or 
52, for four principal reasons. First, subpart M of part 26 would apply 
FFD requirements in a risk-informed manner commensurate with the 
radiological consequences presented by facilities licensed under part 
53. This regulatory strategy is consistent with the current part 26, 
which provides a comprehensive set of deterministic requirements for 
licensees and other entities at facilities that are operating. This 
approach is also consistent with the current subpart K of part 26, 
which provides a more flexible framework for nuclear power reactors 
under construction, where the probabilities of serious radiological 
accidents are lower and consequences from such accidents are less 
severe than at operating plants.
    Second, subpart M of part 26 would enable a part 53 licensee or 
other entity to implement innovative drug testing technologies and 
behavior observation techniques while continuing to demonstrate 
compliance with the part 26 performance objective in Sec.  26.23(b) of 
providing reasonable assurance that individuals are not under the 
influence of any substance or mentally or physically impaired from any 
cause, which in any way adversely affects their ability to safely and 
competently perform assigned duties. These technologies include drug 
and alcohol testing using oral fluid, urine, and hair specimens; 
screening using point of collection testing and assessment (POCTA) 
devices; and monitoring using passive drug and alcohol detection 
instrumentation. Part of the basis to enable the use of innovative drug 
and alcohol testing technologies is to maintain FFD program 
effectiveness should the staff size at a part 53 commercial nuclear 
plant be small and challenge the effective implementation of the 
behavioral observation and drug and alcohol testing programs. Also, a 
commercial nuclear plant that is sited at a geographically remote 
location may present additional challenges to behavioral observation 
and drug and alcohol testing that are not presented by traditional LWR 
facilities licensed under part 50 or 52, such as: efficiency of postal 
services for shipping and controlling biological specimens; proximity 
to drug and alcohol collection facilities that are reasonably 
equivalent to that described in subpart E of part 26; availability of 
internet and cellular services to enable same-time discussions among 
the Medical Review Officer (MRO), donor, and laboratory; accessibility 
to substance abuse treatment services described in subpart H of part 
26; and proximity to an MRO (or management and clinical staff) to 
evaluate potential impairment caused by fatigue and/or substance use or 
abuse, for-cause and post-event occurrences, and the individual's 
potential to return to duty.
    A part 53 commercial nuclear plant that is sited in a 
geographically remote location and has a small staff size may present 
implementation challenges and the potential for small group dynamics to 
impact FFD program effectiveness. Particularly in isolated 
environments, psychological phenomena known as ``groupthink'' may take 
effect and could impact the effectiveness of BOPs and the ability to 
effectively manage safety culture. For example, in circumstances where 
small staffs are drawn from the same small town and thereby have a 
potentially narrow experience base, it could be challenging to maintain 
a safety conscious work environment in which personnel feel free to 
raise safety concerns without fear of retaliation, intimidation, 
harassment, or discrimination, and organizations may resultingly 
experience groupthink-like effects. Groupthink is particularly 
prevalent among cohesive and insulated groups that experience high 
levels of decisional stress.\6\ Small staffs at part 53 commercial 
nuclear plants may therefore be more susceptible to groupthink if they 
are working in an isolated environment where decision-making pressures 
may be high.
---------------------------------------------------------------------------

    \6\ See e.g., Irene W[aelig]r[oslash], Ragnar Rosness, and Stine 
Skaufel Kilska, ``Human performance and safety in Arctic 
environments,'' SINTEF (2018).
---------------------------------------------------------------------------

    Groupthink could have adverse effects on workplace safety culture, 
as studies show that individuals will be more hesitant to speak out 
against practices they deem unsafe for fear of deviating from group 
norms.\7\ Individuals may also be unaware of systematic biases in the 
group decision-making process and may then be less likely to scrutinize 
the potential risks of the group's decision or sufficiently contemplate 
alternative paths of action.\8\ Furthermore, the literature indicates 
that groups make riskier decisions than individuals acting alone due to 
the diffusion of responsibility among group members.\9\

[[Page 86955]]

This phenomenon, known as ``the risky shift,'' also runs counter to a 
safety culture. Accordingly, ``groupthink'' and ``the risky shift'' may 
lead to group behaviors that render behavioral observation less 
effective. As such, alternative approaches to behavior observation 
programs, such as the utilization of video-based surveillance by 
individuals separate from the onsite work unit, could serve to mitigate 
potential issues associated with groupthink. The incorporation of 
remote observation, performed by individuals physically separate from 
the site, could help to bring in independent and objective perspectives 
and help to break patterns of thought and communication that may result 
in groupthink.
---------------------------------------------------------------------------

    \7\ See e.g., Russell Mannion and Carl Thompson, ``Systematic 
biases in group decision-making: implications for patient safety,'' 
International Journal for Quality I Health Care, Vol. 26, No. 6 
(2014): 606-612 (arguing that small group dynamics in healthcare 
teams produce systematic biases in group decision-making because 
healthcare professionals may be reticent to vocalize concerns they 
have about quality of care).
    \8\ See e.g., W[aelig]r[oslash], Rosness, and Kilska (arguing 
that groupthink leads teams to ``develop shared rationalizations 
that bolster a proposed choice, rather than examining alternative 
options and identifying the risks associated with the proposed 
choice''). See also David Hofmann and Adam Stetzer, ``A Cross-Level 
Investigation of Factors Influencing Unsafe Behaviors and 
Accidents,'' Personnel Psychology, Vol. 49 (1996) (finding that in a 
study of fatal accidents involving offshore oil rigs, in the absence 
of standard operating procedures, workers ``equated normal work 
methods (i.e. what everyone else does) with safe and/or ideal work 
methods,'' revealing that the groupthink phenomena will further 
cement modes of work that do not reflect safety protocols in small 
groups that lack strong norms around workplace safety and tacitly 
reward short-cuts that prioritize efficiency over safety).
    \9\ Mannion and Thompson, ``Systematic biases in group decision-
making: implications for patient safety,'' International Journal for 
Quality I Health Care, Vol. 26, No. 6 (2014): 606-612.
---------------------------------------------------------------------------

    Even without the influence of small group dynamics, there are other 
practical constraints to implementing FFD requirements, such as random 
drug and alcohol testing, among small staffs. Random testing is less 
effective when applied to small staff sizes because it may be easier 
for staff to communicate and predict when individuals will be subject 
to drug and alcohol testing. Furthermore, if a facility is sited in a 
remote location, program implementation could be challenged by the 
following factors: limited mail services to laboratories certified by 
the U.S. Department of Health and Human Services (HHS), availability of 
local clinical or medical options for treatment and determinations of 
fitness by an MRO or Substance Abuse Expert, and use of offsite drug 
and alcohol collection facilities.
    The increased potential for small staff sizes to impact FFD policy 
compliance warrants an approach to FFD that emphasizes performance over 
prescriptive requirements that may be ineffective or infeasible at 
these facilities. Therefore, the NRC proposes the subpart M of part 26 
framework to provide a performance-based approach to FFD. For example, 
proposed Sec.  26.603(d) would use existing part 26 auditing 
requirements and the reporting requirement in Sec.  26.717, ``Fitness-
for-duty program performance data,'' and clarify how FFD performance 
data would be used to maintain or improve, if necessary, FFD program 
effectiveness. Specifically, Sec.  26.603(d) would require each 
licensee and other entity that elects to implement subpart M of part 26 
to monitor and assess their site-specific performance against the 
preceding year's site performance, the licensee's most recent fleet-
level performance, and the most recent industry performance. Licensees 
and other entities would use these datapoints to develop performance 
measures, which would be qualitative descriptions of the specific FFD 
program elements, and threshold values for each performance measure 
that, if exceeded, would indicate a performance deficiency. Each 
licensee and other entity would compare its site's current performance 
data against the performance measures and, if a threshold is exceeded, 
the licensee or other entity would be required to take corrective 
actions to restore performance. Also, the NRC proposes a change control 
requirement to allow a licensee or other entity to change its subpart M 
of part 26 FFD program while ensuring that FFD program effectiveness is 
maintained.
    Lastly, subpart M of part 26 would consolidate the applicable FFD 
requirements by placing in one subpart all proposed part 26 
requirements (either new requirements or cross-references to existing 
part 26 requirements) for part 53 licensees and other entities. This 
should help licensees and other entities implement the requirements 
because it would enable easy cross-reference to similar requirements in 
other subparts that are being implemented by non-part 53 licensees and 
entities subject to part 26. Understanding how other licensees or other 
entities implement similar FFD requirements may facilitate the sharing 
of operating experience in program implementation.
    The use of innovative technologies and a risk-informed performance-
based framework parallels the considerations presented in the Advanced 
Reactor Policy Statement. As stated in the policy statement, 
``[S]implified systems should facilitate operator comprehension, 
reliable system function, and more straightforward engineering 
analysis.'' Furthermore, these same attributes may reduce potential 
radiation exposures, help prevent the theft of nuclear materials, and 
use technology and design innovations. Should these components and 
systems be designed, implemented, and maintained to minimize reliance 
on human actions and leverage technology and innovation, then the 
robust and prescriptive FFD requirements in, for example, subparts B, 
``Program Elements,'' and E of part 26 could be scaled to the part 53-
licensed facility and its operation. This strategy would be implemented 
in the subpart M of part 26 framework.
    Even though current subpart K of part 26, provides for FFD 
requirements commensurate with the radiological consequences presented 
by a nuclear power plant construction site, proposed subpart M of part 
26 would not allow part 53 licensees and other entities to implement 
the requirements in subpart K. The principal reasons are that (without 
significant changes to subpart K that would be outside the scope of 
this rulemaking): (1) subpart K does not apply to holders of MLs who 
assemble or test a reactor; (2) subpart K only applies during 
construction, whereas subpart M would apply during construction, 
operation, and decommissioning through implementation of the insider 
mitigation program (IMP) required by Sec.  73.55 or proposed Sec.  
73.100; (3) subpart K does not address training, authorization as 
defined in Sec.  26.5, and MRO performance; (4) subpart K does not 
expressly authorize the use of innovative drug and alcohol testing 
technologies; (5) subpart K does not describe the use of time-dependent 
alcohol limits or special analysis testing of dilute urine specimens; 
and (6) subpart K has less rigor in the protection of worker rights and 
sensitive information than that proposed in subpart M.
    Despite the differences between subparts K and M of part 26, the 
requirements in subpart M would be essentially equivalent to many in 
subpart K that were implemented by the licensees of Vogtle Nuclear 
Station and V.C. Summer Nuclear Station when they were constructing 
four commercial nuclear power reactors and NRC inspection and operating 
experience evaluation determined that the use of subpart K contributed 
to adequately protecting the public health and safety and the common 
defense and security. Further, given the risk profile posed by 
facilities licensed under part 53 and the proposed additional 
requirements in subpart M of part 26 that were developed from operating 
experience and other part 26 subparts (but are not included in subpart 
K of part 26), the NRC concludes that if licensees and other entities 
effectively implement the proposed requirements in subpart M of part 
26, then individuals subject to the rule should be fit for duty and 
trustworthy and reliable.

B. Proposed Changes to Part 26, Subparts A Through E and I

    Section 26.3(d) is the applicability paragraph for contractor/
vendors (C/Vs) who implement FFD programs or program elements, to the 
extent that the licensees and other entities specified in Sec.  26.3(a) 
through (c) rely on those C/V FFD programs or program elements to meet 
the requirements of part 26. Section 26.3(d) would be amended to

[[Page 86956]]

address part 53 licensees and other entities in proposed Sec.  26.3(f).
    Proposed Sec.  26.3(f) would place part 53 licensees or other 
entities within the scope of part 26. For licensees and other entities 
of a part 53 commercial nuclear plant, except a holder of an ML, the 
FFD program would be required to be implemented no later than the start 
of construction activities. The holder of an ML would need to implement 
its FFD program before commencing activities that assemble a reactor.
    Current Sec.  26.4 describes FFD program applicability to 
categories of individuals. These categories are based on the duties, 
responsibilities, and the types of access an individual may possess. 
The NRC proposes to amend Sec.  26.4 to include licensees and other 
entities described in Sec.  26.3(f). The NRC expects that not all 
categories of individuals described in current Sec.  26.4 would be 
applicable to all part 53 facilities. The NRC is proposing regulatory 
guidance in DG-5073, ``Fitness-of-Duty Programs for Commercial Nuclear 
Plants and Manufacturing Facilities Licensed Under 10 CFR part 53,'' 
and DG-5078, ``Fatigue Management for Nuclear Power Plant Personnel at 
Commercial Nuclear Plants Licensed Under 10 CFR part 53,'' to help 
address program applicability to certain individuals.
    Section 26.4(a)(1) and (a)(4) would be amended to account for the 
possibility that certain individuals may perform or direct the 
performance of operational and maintenance activities from a remote 
facility (for example, a remote-control station) for licensees or other 
entities licensed under part 53.
    The framework of the current part 26 does not account for 
individuals who perform operating and maintenance duties at remote 
facilities. Although current Sec.  26.4(a)(1) does not limit the 
operating of applicable SSCs to onsite operating, Sec.  26.5 limits the 
definition of ``Maintenance,'' for the purposes of Sec.  26.4(a)(4), to 
include only ``onsite maintenance activities.'' In the 2008 part 26 
final rule preamble, the NRC explained that the work hour requirements 
apply to those individuals who perform maintenance activities within 
the licensee's owner-controlled area. Furthermore, regarding the 
direction of applicable operations and maintenance activities, current 
Sec.  26.4(a)(1) and (4) address only individuals who perform ``onsite 
direction.''
    Under the proposed amendments to part 26, the limitation of 
``onsite'' activities to those performed within the owner-controlled 
area would still apply to facilities licensed under part 50 or 52. 
However, for licensees and other entities described in Sec.  26.3(f), 
the NRC would remove the ``onsite'' limitation to include activities 
performed both within the owner-controlled area as well as operations 
and maintenance duties performed at remote facilities where safety-
significant systems and components are expected to be operated within 
the design basis of the commercial nuclear plant.
    In the 2008 part 26 final rule, the purpose of limiting 
``directing'' activities to those ``directing'' activities that are 
conducted onsite was to avoid requiring work hour controls for 
individuals performing incidental duties, consistent with Sec.  
26.205(b)(5), from an offsite location in instances where those duties 
might be considered to be ``directive'' in nature. Under the proposed 
amendments to part 26, the exclusion of incidental duties while 
calculating work hours would still be applicable for licensees and 
other entities licensed under part 53. However, for these licensees and 
other entities, beyond instances of incidental duties, the direction of 
operations and maintenance activities associated with safety-
significant SSCs, when performed at remote facilities, would be 
considered in an equivalent fashion as direction performed at non-
remote facilities, for the purposes of administering work hour 
controls.
    Proposed Sec.  26.4(b) would include in an FFD program individuals 
who are granted unescorted access to the protected area of a facility 
licensed under part 53 and do not perform or direct the performance of 
the duties described in Sec.  26.4(a). This requirement would 
contribute to the defense-in-depth regulatory framework that helps 
provide that individuals who have unescorted access are fit for duty, 
trustworthy, and reliable. For example, the NRC is proposing amendments 
to part 73 to require a part 53 licensee to subject individuals to a 
series of reviews to help determine whether those individuals are 
trustworthy and reliable before granting them unescorted access to the 
facility's protected area.
    The NRC would amend Sec.  26.4(c) to include in an FFD program 
individuals who are assigned to physically report to the part 53 
licensee's emergency response facility (or facilities) or participate 
remotely in emergency response activities, and individuals without 
unescorted access to the part 53 facility who, remotely or otherwise, 
make decisions and/or direct actions regarding plant safety or 
security. Part 53 commercial nuclear plants may be licensed for and 
rely upon offsite facilities to fulfill the role of a Technical Support 
Center or Emergency Operations Facility. Therefore, the proposed rule 
would account for such offsite facilities or remotely performed 
activities. Further, the use of personnel to operate systems and 
components, maintain and surveil SSCs, and respond to plant conditions 
and security events may be different than those included in the 
Technical Support Center or Emergency Operations Facility team for 
power reactors currently licensed under part 50 or part 52.
    For the individuals whose duties for the licensees and other 
entities in Sec.  26.3(c) require the individuals to have the types of 
access or perform the activities listed in Sec.  26.4(e)(1) through (6) 
at the location where the commercial nuclear plant will be constructed 
and operated, current Sec.  26.4(e) requires them to be subject to an 
FFD program that satisfies all the requirements of part 26 except 
subparts I and K. The NRC would amend Sec.  26.4(e) to except subpart M 
as well as subparts I and K. The NRC would also amend Sec.  26.4(e) to 
include in an FFD program the individuals whose duties for the 
licensees and other entities in Sec.  26.3(f) require the individuals 
to have the types of access or perform the activities listed in Sec.  
26.4(e)(1) through (6) or perform construction activities as defined in 
Sec.  26.5.
    Section 26.4(e)(4) would be revised to include in an FFD program 
individuals who witness or determine inspections, tests, and analyses 
certifications required under part 53 because current Sec.  26.4(e)(4) 
includes the individuals who perform the same duties under part 52.
    The proposed rule would amend Sec.  26.4(f) to require individuals 
who construct or direct the construction of safety- or security-related 
SSCs at facilities licensed under part 53 to be subject to an FFD 
program under subpart M of part 26 or an FFD program that demonstrates 
compliance with all of the requirements of part 26 except for subparts 
I, K, and M of part 26.
    Section 26.4(g) is the applicability paragraph for FFD program 
personnel (e.g., the FFD manager, MRO, and technicians) and persons who 
perform AA determinations (e.g., the licensee- or other entity-
designated Reviewing Official). This section would be amended to 
address part 53 licensed facilities. Specifically, a part 53 licensee 
or other entity would use FFD program personnel to implement its FFD 
program as well as other assigned individuals who are not involved in 
the day-to-day operations of the program to implement specific elements 
of its FFD program, such as the collection of a

[[Page 86957]]

specimen for drug or alcohol testing. These individuals would be held 
accountable for program implementation, including consistent 
implementation of protections afforded to all individuals subject to 
the FFD program.
    Section 26.4(h) would be amended to include subpart M of part 26.
    The NRC proposes to include several new definitions in Sec.  26.5, 
``Definitions,'' and amend some existing definitions. The NRC is 
proposing to add a definition for ``Biological marker.'' The proposed 
definition would be consistent with ``Biomarker'' defined by the HHS in 
its Mandatory Guidelines for Federal Workplace Drug Testing (HHS 
Guidelines) using oral fluid as the biological specimen to be tested 
(84 FR 57554; October 25, 2019). However, the proposed definition for 
Sec.  26.5 would add that the endogenous substance used to validate 
that the biological specimen ``was produced by the donor'' because 
subpart M of part 26 proposes to have the MRO evaluate any discrepant 
biological marker identified in a biological specimen collected from a 
donor.
    The NRC is proposing a definition for the word ``Change'' as used 
in the proposed Sec.  26.603(e), ``FFD program change control,'' 
process. The proposed definition would be consistent with the 
definition of ``Change'' for a part 50 or 52 licensee's emergency plans 
in Sec.  50.54(q)(1)(i).
    The NRC proposes to revise the definition of ``Constructing or 
construction activities'' to clarify that for licensees or other 
entities in Sec.  26.3(f), the definition of ``Construction'' would be 
that as proposed in Sec.  53.020.
    The definitions of ``Contractor/vendor'' (C/V) and ``Other entity'' 
would be revised to make them applicable to part 53 licensees. A holder 
of an ML under part 53 could be a C/V under the proposed C/V 
definition.
    The NRC is proposing a definition for ``Illicit substance'' because 
this phrase is used in subpart M of part 26 and would address 
substances that cause impairment and possible addiction but are not an 
``illegal drug'' as defined in Sec.  26.5. This proposal is based on 
operating experience where individuals have admitted to using common 
household, non-drug substances to achieve a high or satisfy an 
addiction. These common household items include, but are not limited to 
nitrous oxide, butane, propane, glue, paint vapors, lighter fluid, nail 
polish remover, degreasers, permanent markers, and methyl alcohol 
(which is found in hand sanitizer and mouthwash).
    The definition of ``Questionable validity'' would be revised to 
make it applicable to an FFD program implemented under subpart M of 
part 26, which would include all biological specimens.
    The NRC is proposing a definition for ``Reduction in FFD program 
effectiveness'' because this phrase, similar to the proposed definition 
for ``Change,'' is used in proposed Sec.  26.603(e). The proposed 
definition is generally consistent with the definition of ``Reduction 
in effectiveness'' provided for emergency plans in Sec.  
50.54(q)(1)(iv).
    The proposed rule would make the current definition of ``Reviewing 
official'' applicable to those licenses and other entities in Sec.  
26.3(f).
    The current part 26 definition of ``Safety-related structures, 
systems, and components'' would be amended to use the NRC's proposed 
definition in Sec.  53.020 for the part 53 licensees and other entities 
described in Sec.  26.3(d) and (f).
    The NRC would amend the definition of ``Security-related SSCs'' in 
Sec.  26.5 to make it applicable to a licensee or other entity 
described in Sec.  26.3(d) and (f).
    The NRC proposes a definition for ``Special Nuclear Material'' that 
would refer to the definition in Sec.  70.4, ``Definitions,'' of part 
70 to ensure consistency.
    The NRC is proposing a revision of the definition of ``Unit 
outage'' to account for the potential use of commercial nuclear plants 
for purposes other than electricity generation.
    Section 26.21, an applicability statement for part 26 FFD programs, 
would be amended to include licensees and other entities described in 
Sec.  26.3(f) that choose to implement an FFD program that implements 
all part 26 requirements, except those in subparts K and M of part 26.
    Section 26.51, ``Applicability,'' would be amended to apply to 
licensees and other entities described Sec.  26.3(f) that elect not to 
implement the requirements in subpart M of part 26 for the categories 
of individuals in Sec.  26.4 and those licensees and other entities 
that elect to implement the requirements in Sec.  26.605.
    Section 26.53(e), (e)(1) and (3), and (g) through (i), which are 
general provisions for granting and maintaining authorization, would be 
amended to apply to licensees and other entities described Sec.  
26.3(f).
    Section 26.63(d), a suitable inquiry requirement, would be amended 
to apply to licensees and other entities described Sec.  26.3(f).
    Section 26.73, the applicability statement for subpart D of part 
26, would be amended to apply to licensees and other entities described 
Sec.  26.3(f) that elect not to implement the requirements in subpart M 
of part 26 for the categories of individuals in Sec.  26.4 and those 
licensees and other entities that elect to implement the requirements 
in Sec.  26.605(b).
    Section 26.81, the purpose and applicability statement for subpart 
E of part 26, would be amended to apply to licensees and other entities 
described in Sec.  26.3(f) that elect not to implement the requirements 
in subpart M of part 26 for the categories of individuals in Sec.  26.4 
and those licensees and other entities that implement proposed Sec.  
26.605(a) or (b). The subpart E requirements to be implemented are 
listed in proposed Sec.  26.607(c)(2)(i) and (c)(2)(ii) and (c)(3).
    Section 26.201, the applicability statement for subpart I of part 
26 would be amended to apply to licensees and other entities described 
in Sec.  26.3(f). Also, the applicability statement would be divided 
into two paragraphs for clarity.
    The NRC proposes to add Sec.  26.202, ``General provisions for 
facilities licensed under part 53,'' for licensees or other entities 
described in proposed Sec.  26.3(f) that elect to implement the 
requirements in subpart I of part 26 in accordance with Sec.  26.604 
and Sec.  26.605. Section 26.202 would establish requirements 
equivalent to those in current Sec.  26.203, ``General provisions,'' 
which is applicable to part 50 and 52 licensees. The NRC would add the 
separate Sec.  26.202 because Sec.  26.203 refers to various 
requirements under subpart B of part 26, which would not be applicable 
to facilities licensed under part 53 that implement subpart M of part 
26.
    Additionally, Sec.  26.202(c), ``Training and assessments,'' unlike 
Sec.  26.203(c), ``Training and examinations,'' would not include a 
comprehensive examination requirement because trainee assessment is 
conducted as part of a SAT that would be required as proposed under the 
FFD program training requirements in Sec.  26.608.
    Proposed changes in Sec. Sec.  26.205, 26.207, and 26.211 would add 
references to new requirements in subparts I and M of part 26 that 
would be applicable specifically to licensees and other entities in 
Sec.  26.3(f). The NRC would not change the specific provisions for 
work hour requirements in current Sec.  26.205(d). However, as 
addressed in the discussion of proposed changes to Sec.  26.4(a), 
whether a licensee or other entity under part 26 would need to 
implement work hour controls

[[Page 86958]]

for certain individuals or groups would be dependent, in part, on 
determinations reached by that licensee's risk-informed evaluation 
process.
    Proposed changes to Sec. Sec.  26.207(a)(1)(ii) and 26.211(b) would 
allow licensees and other entities in Sec.  26.3(f) to perform face-to-
face assessments to support the approval of work hour control waivers 
and the conduct of fatigue assessments, respectively, using electronic 
communications. These proposals would allow supervisors to conduct such 
assessments from a remote location under appropriate circumstances. 
Such remotely conducted assessments would need to be supported by 
someone who is present in-person with the individual being assessed and 
who is trained in accordance with the requirements of either Sec.  
26.29 and Sec.  26.203(c) or Sec.  26.608 and Sec.  26.202(c). The 
reasoning for these proposals and the associated need for in-person 
support to augment electronic communications is addressed further in 
the preamble discussion of proposed Sec.  26.619.

C. Proposed Requirements for Part 26, Subpart M

    The proposed rule would add a new subpart M to part 26 that would 
provide alternative FFD requirements for part 53 licensees and other 
entities.
    Proposed Sec.  26.601 would make subpart M of part 26 applicable to 
part 53 licensees and other entities, at their discretion. If a 
licensee or other entity in Sec.  26.3(f) does not elect to implement 
an FFD program that demonstrates compliance with the requirements of 
subpart M, then the individuals specified in Sec.  26.4 would be 
subject to an FFD program that demonstrates compliance with all part 26 
requirements, except for those requirements in subparts K and M.
    Proposed Sec.  26.603(a) would require an applicant to provide a 
description of its FFD program and its implementation within its 
application for a license. This requirement is equivalent to the 
existing requirements in Sec. Sec.  26.401(b) and 52.79(a)(44). The 
entities that would be required to submit these FFD program 
descriptions are certain applicants that would comply with the part 53 
application requirements in subpart H. In subpart H, Sec.  
53.1309(a)(6) would require an applicant for a CP to provide a 
description of its FFD program in its PSAR. Under Sec. Sec.  
53.1279(b)(4), 53.1369(x), and 53.1416(a)(24), an applicant for an ML, 
OL, and COL, respectively, would be required to provide a description 
of its FFD program in its FSAR.
    Unlike an application for a license, a description of an FFD 
program does not receive NRC review for possible approval. The 
applicant provides the NRC with information about the applicant's 
proposed FFD program to inform the NRC's inspection program and to 
demonstrate that the FFD program will be effectively implemented before 
a licensee or other entity commences any activity making individuals at 
the NRC-licensed facility subject to the FFD program.
    Proposed Sec.  26.603(a)(1) would require a summary description of 
the analysis described in Sec.  26.603(c), if performed. The analysis 
should describe the operation of the facility. This would include 
informing the Commission of: (1) the principal individuals assigned by 
job title (work category) and a summary description of the human 
actions (e.g., monitoring, operating, responding, surveillance, 
oversight, etc.) that they perform to maintain the facility in a safe 
operating or shutdown condition; (2) the principal individuals by job 
title and a summarized description of the human actions to secure and 
protect the facility (without providing sensitive information); (3) the 
estimated total population of individuals subject to the FFD program 
and per shift by job description; and (4) references to supporting 
documentation. The purpose of these descriptions is to enable an NRC 
assessment of the licensee's or other entity's analysis and the 
required human actions to operate, monitor, surveil, maintain, and 
secure the facility within its design and licensing basis so that if an 
operational or security-related event were to occur, the facility would 
respond as designed and licensed and the calculated radiological dose 
consequences would not exceed the consequences described in Sec.  
53.860(a)(2). This is important because facilities that implement Sec.  
26.604 are expected to have very small staff sizes and may be sited in 
geographically remote locations, both of which could challenge 
effective implementation of the FFD program.
    Proposed Sec.  26.603(a)(2) would require the applicant to state 
what FFD program it plans to implement.
    Proposed Sec.  26.603(a)(3) would require a discussion that informs 
the NRC of the applicability of the applicant's FFD program to 
individuals who perform safety- or security-significant activities. 
This description should summarize any key differences between the staff 
at the site and any remote facility and the categories of individuals 
in Sec.  26.4. The principal purpose of providing this description 
would be to inform the NRC of any substantial differences in the 
applicability of the FFD program to the categories of individuals in 
Sec.  26.4.
    Proposed Sec.  26.603(a)(4) would require a description of the drug 
and alcohol testing and fitness determination process to be implemented 
through the licensee's or other entity's procedures, including the 
collection and testing facilities to be used, biological specimens to 
be collected, and sanctions to be imposed upon a confirmed FFD policy 
violation. This process includes how individuals who test positive for 
a drug or alcohol will be evaluated before being afforded unescorted 
access to the protected area to perform or direct those duties or 
responsibilities making them subject to the FFD program. The principal 
purpose of describing this return-to-duty process is to inform the NRC 
of the behavioral observation strategy (for those facilities that 
implement Sec.  26.604) and/or drug screening and testing strategy.
    Proposed Sec.  26.603(a)(5) would require a summary description of 
the applicant's planned PMRP. This description must provide the 
performance measures and thresholds that the applicant intends to use.
    Proposed Sec.  26.603(b) would establish when the FFD program must 
be implemented and the longevity of the FFD program. This proposal is 
equivalent to the current Sec.  26.3, which states, in part, when 
licensees and other entities must begin implementing their FFD 
programs. Unlike the current part 26 regulations, proposed Sec.  
26.603(b) would expressly state that an FFD program would not be 
applicable during decommissioning of a part 53 facility for licensees 
and other entities specified in Sec.  26.3(f). However, licensees of 
facilities licensed to operate a reactor should be aware that the 
physical protection program under Sec.  73.55, ``Requirements for 
physical protection of licensed activities in nuclear power reactors 
against radiological sabotage,'' and under proposed Sec.  73.100 
include a requirement for the implementation of an IMP, even during 
decommissioning.
    Proposed Sec.  26.603(b) would also require the holder of an ML to 
implement its FFD program no later than the start of activities that 
assemble a reactor. The holder of the ML should establish in its 
procedures when reactor assembly commences and what constitutes 
assembly. For example, the FFD program would not need to be implemented 
for the receipt, storage, inspection, and staging of components and 
systems used to assemble (i.e., build or fabricate) the reactor because 
this is not a current requirement for LWR facilities licensed under 
part 50 or 52. Furthermore, the NRC currently does

[[Page 86959]]

not require that an FFD program be applied to the assembly or 
manufacturing of components (or basic components as defined in Sec.  
21.3), or systems that were fabricated or assembled outside the 
footprint of a commercial power reactor, and this regulatory position 
would also apply to a manufacturing facility.
    Proposed Sec.  26.603(c) would require the applicant, licensee, or 
other entity seeking to implement an FFD program under Sec.  26.604 to 
perform a site-specific analysis to determine whether the facility and 
its operation satisfy the criterion in Sec.  53.860(a)(2). If the 
analysis is performed and demonstrates that the radiological 
consequences presented by the facility and its operation satisfy the 
criterion, then the licensee or other entity could implement the FFD 
program detailed in Sec.  26.604. If the analysis does not demonstrate 
that the facility and its operation satisfy the criterion, then the 
licensee or other entity must implement the FFD program described in 
either Sec.  26.605 or subparts A through I, N, and O of part 26.
    Proposed Sec.  26.603(c) would also require licensees and other 
entities that implement proposed Sec.  26.604 to update the technical 
analysis used to justify compliance with the criterion in Sec.  
53.860(a)(2). This analysis would be updated to reflect changes made to 
the staffing, FFD programs, or offsite support resources described in 
the analysis to show that the facility and its operation continue to 
satisfy the criterion. This is important because facility, operation, 
or staffing changes outside FFD program implementation (e.g., changes 
in the facility safety analysis, physical protection strategies, or the 
security plan, implementing procedures, or contingency response 
strategies) could adversely impact the licensee's or other entity's 
documented analysis demonstrating that the facility and its operation 
satisfy the criterion if event sequences require human action.
    Proposed Sec.  26.603(d) would require the establishment of a PMRP. 
The concept of a PMRP is not new. This requirement would consolidate 
for part 53 the requirements in current Sec. Sec.  26.41, ``Audits and 
corrective actions''; 26.415, ``Audits''; 26.717, ``Fitness-for-duty 
program performance data''; and 26.183(c), which describes MRO 
responsibilities. The proposal would state that the licensee or other 
entity must monitor the effectiveness of its FFD program by comparing 
performance data against performance measures and thresholds. The 
development of quantitative thresholds would be new, but this is born 
from licensees and other entities with facilities licensed under parts 
50 or 52 already collecting, reviewing, and reporting FFD performance 
data. Additionally, the benefit of quantitatively measuring FFD program 
performance against established thresholds benefits a licensee's and 
other entity's determination of whether they are maintaining FFD 
program performance in a manner that demonstrates compliance with the 
performance objectives in Sec.  26.23.
    The NRC is proposing the PMRP because the subpart M of part 26 
requirements would enable a high degree of flexibility in FFD program 
implementation (e.g., drug testing). A licensee or other entity would 
not only have options in the type of FFD program they may implement 
under part 26, but they would have options in the types of biological 
specimens they may test for drugs, where to collect the biological 
specimens (e.g., at the NRC-licensed facility or offsite at a local 
hospital or clinic), and the use of collection and assessment devices 
to screen individuals for drugs and alcohol. These FFD program 
flexibilities could cause FFD programs under subpart M of part 26 to 
become very site-specific, necessitating performance measures to enable 
the licensee or other entity to maintain the effectiveness of its FFD 
program.
    Fitness-for-duty program effectiveness would be determined by 
comparing actual performance against the performance measures and 
thresholds. The result of that comparison would inform licensee or 
other entity decisions whether to change FFD program elements to 
address a performance deficiency. Also, the thresholds would have 
sufficient margin, based on operating experience, before conditions 
adverse to safety and security may occur should an individual be 
identified as impaired or not trustworthy and reliable. The potential 
of a human-related failure causing a condition adverse to safety and 
security is dependent on the duties and responsibilities of the 
individual and the defense-in-depth designed to prevent or mitigate an 
adverse consequence. The PMRP would account for this by requiring the 
review of FFD performance data, in part, by work category, C/V, and 
individuals employed by the licensee who are not a C/V as defined in 
Sec.  26.5 (i.e., a licensee employee).
    Proposed Sec.  26.603(d)(1) would require the licensee or other 
entity to document and maintain its PMRP. Proposed Sec.  
26.603(d)(1)(i) would require that the performance measures be 
identified and designed to monitor FFD program performance. Proposed 
Sec.  26.603(d)(1)(i)(A) would require the FFD program of a licensee or 
other entity subject to the requirements of Sec.  26.604 to include 
monitoring of the BOP. The purpose of this monitoring is to help ensure 
that individuals subject to the FFD program are observing the behaviors 
of others, are being observed themselves, and are reporting FFD 
concerns to licensee- or other entity-designated individuals. The other 
performance measures would include occurrence of FFD policy violations 
evaluated by licensee employee, C/V, and labor category, and occurrence 
of individuals with potentially disqualifying information or who 
possessed an FFD prohibited item.
    Proposed Sec.  26.603(d)(1)(i)(B) would require the FFD program of 
a licensee or other entity that is either subject to the requirements 
of Sec.  26.604 and has implemented a drug testing program at its 
discretion, or is subject to the requirements of Sec.  26.605, to 
include the performance measures identified in Sec.  26.603(d)(1)(i)(A) 
and those necessary to monitor the effectiveness of the drug and 
alcohol testing program. The drug and alcohol measures would include 
the monitoring of FFD performance data for pre-access and random 
testing and subversion attempts by the categories of licensee employee, 
C/V, and labor category.
    Proposed Sec.  26.603(d)(1)(ii) would require the licensee or other 
entity to establish thresholds for each performance measure. Initial 
thresholds must be based on FFD performance data from comparable 
facilities subject to part 26, the licensee's or other entity's fleet-
level program performance if applicable, and industry FFD performance 
data. This provision introduces the requirement to ``maintain FFD 
program effectiveness.'' This terminology describes a performance-based 
regulatory strategy in which the licensee or other entity must 
initially establish a level of performance that is representative of 
other facilities in the licensee's fleet of facilities subject to part 
26, if applicable, and the FFD performance of comparable facilities 
subject to part 26.
    Proposed Sec.  26.603(d)(1)(iii) would require that the licensee or 
other entity evaluate FFD data as it is received to determine whether a 
threshold has been exceeded. Historical FFD performance data for the 
current LWR fleet indicates that, for particular work categories and 
employment types, few FFD policy violations occur per year. Therefore, 
for work categories that may be significant to worker safety (e.g., 
radiation protection technicians), physical

[[Page 86960]]

protection (i.e., security personnel), or safety (i.e., NRC-licensed 
operators and individuals who perform or direct the performance of 
activities that a risk-informed evaluation process has shown to be 
significant to public health and safety), a single FFD policy violation 
could be a significant occurrence and warrant corrective actions. Based 
on licensee-submitted FFD-related reports under Sec. Sec.  26.417, 
26.419, 26.717, and 26.719, licensees and other entities with 
facilities licensed under parts 50 or 52 implement some form of 
corrective action that is typically scaled to the significance of the 
violation. These corrective actions have included counseling, follow-up 
drug and/or alcohol testing, remedial training, generic announcements 
to the workforce, and reviews of recently performed or directed work by 
the individual suspected of being impaired. Proposed Sec.  
26.603(d)(1)(iii) would require that the PMRP include a year-to-year 
comparison of FFD performance data to help provide assurance that an 
adverse trend in FFD program performance would be identified if 
occurring. This proposed requirement was developed from the annual FFD 
performance data reporting requirements in Sec. Sec.  26.417(b)(2) and 
26.717. In particular, the proposed year-to-year comparison of FFD 
performance data is equivalent to Sec.  26.717(c), which requires, in 
part, licensees and other entities to analyze their performance data at 
least annually and take appropriate actions to correct any identified 
program weaknesses.
    Proposed Sec.  26.603(d)(1)(iv) would require the licensee or other 
entity to perform and document quantitative and qualitative reviews. 
These reviews would be performed in three program areas: protections 
afforded to individuals subject to the FFD program, laboratory test 
results and MRO performance, and change control. The purpose of these 
reviews would be to specifically target performance within the three 
program areas to assess whether the outcomes resulting from the 
implementation of procedure requirements are contributing to FFD 
program effectiveness. The proposed reviews would not require the 
establishment of measures and thresholds because the reviews are 
expected to result in qualitative findings regarding program 
effectiveness. Qualitative findings and observations could still result 
in the consideration of corrective actions in the targeted program 
areas.
    Proposed Sec.  26.603(d)(1)(iv)(A) would require the licensee or 
other entity to monitor whether its FFD program is affording 
appropriate protections to individuals subject to the FFD program. The 
review of these protections would include, in part, assessing the 
licensee's or other entity's protection of the following: privacy 
during the specimen collection process; specimen integrity, custody, 
and control; information gathered from FFD program implementation; and 
due process during appeals of FFD policy violations.
    Proposed Sec.  26.603(d)(1)(iv)(B) would require, in part, a review 
of laboratory test results and MRO performance. Effective performance 
by the laboratory (e.g., obtaining and communicating accurate test 
results) and MRO (e.g., correct evaluation of the laboratory test 
results based on Sec.  26.185 or HHS Guidelines) would result in three 
significant outcomes: (1) protection of the donor from an inaccurate 
FFD policy violation determination; (2) protection of the donor, other 
individuals, and the facility from potential harm should the donor be 
impaired or not trustworthy and reliable; and (3) a performance-based 
assessment of both the laboratory and MRO. This last outcome could 
facilitate actions to improve laboratory performance, MRO training 
under Sec.  26.607(m), or both. Proposed Sec.  26.603(d)(1)(iv)(B) 
would also require a comparative analysis between the POCTA screening 
result(s) and the corresponding specimen test results obtained from the 
HHS-certified laboratory if the POCTA indicated a positive, 
adulterated, substituted, or invalid screening result or discrepant 
biological marker, to assess the effectiveness of the POCTA and to 
inform MRO decisions under Sec.  26.185 or Sec.  26.607(m)(6). The 
results of this biennial review could also inform the conduct of 
laboratory audits.
    Proposed Sec.  26.603(d)(1)(iv)(C) would require that the change 
control requirement in proposed Sec.  26.603(e) be included in the 
biennial program review to help ensure that changes implemented over 
the life of the facility do not result in a reduction in program 
effectiveness even if a mitigating action was implemented for the 
specific change. This requirement was developed from Sec. Sec.  
26.137(f) and 26.713(d). This part of the review would require an 
assessment of all changes since the last review and their potential 
aggregated impact on FFD program effectiveness. For example, if last 
year the licensee elected to contract with a different MRO and this 
year the licensee implemented a new type of POCTA device, each of those 
program changes probably would not have resulted in a recognizable 
reduction in FFD program effectiveness. But, if the drug testing 
positivity rate (or FFD policy violations) for C/Vs decreased markedly 
during a future maintenance outage that required many C/Vs, then the 
reduction could indicate, for example, that the POCTA device was not as 
effective as determined by a forensic toxicologist review under 
Sec. Sec.  26.603(e) and 26.607(h) or that the new MRO was improperly 
crediting prescription medication for laboratory-confirmed positive 
test results.
    Proposed Sec.  26.603(d)(2) would state when the licensee or other 
entity must implement corrective actions. This requirement would be 
equivalent to the requirement in current Sec.  26.415(b) and was 
developed from requirements contained in Sec. Sec.  26.41(a) and (f), 
26.127(e), 26.129(b)(1)(i), 26.137(f)(3) through (5), 26.155(a)(6), 
26.157(e), 26.159(b)(1)(i), and 26.203(e)(2). Corrective actions must 
be implemented to correct root causes, contributing causes, or both. 
There is margin built into the FFD performance thresholds and 
qualitative factors (e.g., to account for potential changes in drug and 
alcohol testing performance data when there is a large influx of C/Vs 
to perform maintenance) that may influence a licensee or other entity's 
causal determination for an occurrence. Thus, generalized or 
qualitative corrective actions may be implemented like informing 
management and placing a sufficiently descriptive summary of the 
occurrence in a corrective action program for future monitoring to 
assess recurrence.
    However, should the occurrence challenge safety or security or 
significantly exceed a performance threshold even when considering 
qualitative factors and margin, the licensee or other entity should 
implement more robust corrective actions to resolve the cause. An 
example of a challenge to safety or security would be the situation 
when an NRC-licensed operator or maintenance professional had operated, 
surveilled, or maintained safety-significant SSCs and was determined to 
have been impaired by behavioral observation or potentially under the 
influence of a narcotic as determined by an alcohol or drug test or 
screening result. Immediate corrective actions could include, but would 
not be limited to, a licensee or other entity assessment of the duties 
and responsibilities recently performed by the individual. Operating 
experience within the LWR operating reactor community demonstrates few 
FFD policy violations per year per site have been caused by individuals 
who perform or direct the performance of

[[Page 86961]]

safety or security-significant activities. Therefore, any such 
violations of the FFD policy in a particular work category in one year 
could be a significant performance deficiency. These violations could 
be even more significant at part 53 facilities that have a very small 
workforce subject to part 26.
    Proposed Sec.  26.603(d)(3) would require the licensee or other 
entity to biennially assess and document its FFD performance monitoring 
program; this requirement was developed from Sec.  26.41(b). This 
documented review would demonstrate that the performance measures and 
thresholds are appropriate based on site- and licensee's fleet-level 
program performance, if applicable, and industry performance and 
adjusted to maintain FFD program effectiveness. Also, as a result of 
this effort, the licensee or other entity would be in possession of 
lessons learned from fleet-level performance, if applicable, and 
industry performance that could contribute to their own performance 
assessment to maintain program effectiveness.
    Under proposed Sec.  26.603(d)(3)(i), the identified program 
weaknesses and corrective actions resulting from the biennial review 
would be required to be summarized in the licensee's or other entity's 
annual report to the NRC in compliance with either Sec.  26.417(b)(2) 
or Sec.  26.717, as applicable. This information would inform the NRC 
of FFD program weaknesses to facilitate regulatory oversight and enable 
the NRC to aggregate industry data for use in a licensee or other 
entity PMRP.
    Proposed Sec.  26.603(d)(3)(ii) would establish when the biennial 
PMRP review must be completed and when corrective actions from the 
review must be implemented. The NRC selected the May 15th date of odd-
numbered years to help ensure that all FFD programs will maintain their 
previously determined performance measures and thresholds or reset them 
based on FFD program performance early in the year in which the 
biennial review was conducted. This would assist in obtaining quality 
FFD performance data over two annual reporting cycles and evaluating 
whether previous corrective actions were effective.
    In proposed Sec.  26.603(e), the NRC proposes a change control 
requirement for subpart M of part 26 FFD programs. Requiring licensees 
and other entities to demonstrate compliance with certain requirements 
before implementing changes to their FFD programs would be necessary 
for two primary reasons. First, proposed changes to a licensee's or 
other entity's FFD program could affect the analysis performed by the 
licensee or other entity under proposed Sec.  26.603(c), which helps 
determine the FFD program requirements that must be implemented. If 
this analysis changes, then the licensee's or other entity's FFD 
program requirements might change. Second, the requirements in subpart 
M of part 26 are performance based. Therefore, FFD program 
implementation may change periodically in response to societal changes 
in substance abuse or from PMRP implementation. Change control 
therefore relies on the licensee or other entity maintaining its 
procedures in a manner that details how its FFD program is to be 
implemented while incorporating changes, with documentation that 
justifies the changes to support the PMRP, audits, and NRC inspection.
    Proposed Sec.  26.603(e)(1) would permit the licensee or other 
entity to implement changes to its FFD program if it performs and 
retains an analysis demonstrating that the change does not reduce the 
effectiveness of the FFD program or the change was necessitated or 
justified by a change to part 26, laboratory processes, or guidance 
issued by the HHS or NRC. The proposed change control requirement would 
enable flexibility in program implementation should the NRC or HHS 
change its drug testing procedures (as implemented by the licensee or 
other entity through its procedures) in response to changes in societal 
substance abuse or drug testing technologies.
    The proposed change control requirement was developed from the 
change control requirements in Sec.  50.54(p) and (q)--the change 
control requirements for security and emergency plans, respectively. 
However, unlike these two requirements, the NRC does not review and 
approve a licensee's or other entity's FFD program or its implementing 
procedures, and the FFD program is not licensing-basis information as 
described in Sec.  53.1300.
    Proposed Sec.  26.603(e)(2) would require that if a change reduces 
FFD program effectiveness, then the licensee must implement a 
mitigating strategy so the FFD program, as revised, will continue to 
demonstrate compliance with the performance objectives in Sec.  26.23 
and not result in a reduction in program effectiveness.
    Proposed Sec.  26.603(e)(3) would prohibit, with one exception, the 
use of the change control process to reduce the minimum panel of drugs 
to be tested and would reference the drugs listed in proposed Sec.  
26.607(c)(1). Proposed Sec.  26.607(c)(1) would reference current Sec.  
26.31(d)(1), which states that, at a minimum, licensees and other 
entities shall test for marijuana metabolite, cocaine metabolite, 
opioids (codeine, morphine, 6-acetylmorphine, hydrocodone, 
hydromorphone, oxycodone, and oxymorphone), amphetamines (amphetamine, 
methamphetamine, methylenedioxymethamphetamine, and 
methylenedioxyamphetamine), phencyclidine, and alcohol. The testing of 
these drugs and drug metabolites, except phencyclidine, and alcohol is 
necessary for the FFD program to remain effective. Also, there is no 
proposed subpart M of part 26 requirement stating that this panel of 
drugs and drug metabolites needs to consist of only scheduled 
drugs.\10\ This flexibility would account for the situation where an 
impairing substance becomes prevalent in society and a licensee or 
other entity elects to add the substance to their panel of substances 
to be tested prior to it being scheduled by the Drug Enforcement 
Administration.
---------------------------------------------------------------------------

    \10\ The Drug Enforcement Administration classifies drugs, 
substances, and certain chemicals used to make drugs into five (5) 
distinct categories, depending upon the drug's acceptable medical 
use and the drug's abuse or dependency potential. These categories 
appear as Schedules I through V of section 202 of the Controlled 
Substances Act (21 U.S.C. 812). Schedule I drugs have a high 
potential for abuse, have no currently accepted medical uses in 
treatment in the United States, and lack accepted safety for use 
under medical supervision. At the other end of the classification 
scheme, Schedule V drugs have the least potential for abuse among 
the five categories of drugs, have a currently accepted medical use 
in treatment in the United States, and abuse of the drug may lead to 
limited physical dependence or psychological dependence. For more 
information, see https://www.dea.gov/drug-information/drug-scheduling.
---------------------------------------------------------------------------

    The exception in proposed Sec.  26.603(e)(3) would be that, should 
HHS elect to remove phencyclidine from the panel of drugs and drug 
metabolites to be tested, a licensee or other entity could make this 
change in its FFD program without resulting in a reduction in FFD 
program effectiveness. This outcome would be justified based on the 
very infrequent occurrence rate of FFD policy violations due to 
phencyclidine use since 2010. However, if HHS proposes to remove a 
class of drugs from the panel of drugs to be tested that is listed in 
Sec.  26.31(d)(1), except for phencyclidine, then a licensee or other 
entity may not make a similar change to its panel of drugs to be 
tested, because this change would be a reduction in FFD program 
effectiveness even with a mitigative strategy implemented.
    Changes in the HHS panel of drugs and drug metabolites to be tested 
may also shift from one metabolite to a

[[Page 86962]]

different metabolite for the same drug class (e.g., amphetamines, 
opioids) to be tested. Should HHS issue such a change to its panel, 
this would not be expected to result in a reduction in FFD program 
effectiveness because HHS would be targeting a more prevalent or 
effective metabolite in its drug testing program. This situation could 
occur as HHS gathers more operating experience from Federal Government 
implementation of its HHS Guidelines, or data generated by drug testing 
laboratories and federally mandated drug testing programs required by 
Federal agencies such as the NRC and U.S. Departments of 
Transportation, Energy, and Defense.
    Proposed Sec.  26.603(e)(4) would require that change control 
records be maintained for a 5-year record retention period based on the 
current NRC practice to conduct triennial inspections of licensees' and 
other entities' FFD programs. This would afford the NRC an opportunity 
to review the licensee's or other entity's determination that FFD 
program changes have not reduced the effectiveness of their FFD 
program. Licensees and other entities would also be required to 
summarize each change made under proposed Sec.  26.603(e) in their 
annual FFD performance reports required by Sec.  26.617(b)(2) or Sec.  
26.717, as applicable.
    Proposed Sec.  26.604 would establish the minimum set of FFD 
program requirements for licensees and other entities who have a 
documented analysis that demonstrates that the facility and its 
operation satisfy the criterion in Sec.  53.860(a)(2). For these 
licensees, compliance with the performance objectives in Sec.  26.23 
would be ensured through the BOP; defense-in-depth measures proposed in 
subpart M of part 26 like the PMRP, change control, and audits; and 
other requirements, such as those for AA, physical protection, and 
licensed operators. The adequacy of these measures in satisfying the 
performance objectives is supported by operating experience, which 
demonstrates margin between an FFD-related occurrence and a condition 
adverse to safety or security, as illustrated by for-cause, post-event, 
and random testing data. A facility that satisfies the criterion in 
proposed Sec.  53.860(a)(2) would present a smaller potential 
radiological consequence than a facility that does not satisfy the 
criterion, so the requirements in proposed Sec.  26.604 are scaled to 
the lower risk presented consistent with the Commission's Advanced 
Reactor Policy Statement.
    The disadvantages of implementing the FFD program described in 
proposed Sec.  26.604 would be few. Since drug and alcohol testing 
would not be required, behavioral observation would be the keystone 
requirement in this performance-based framework to provide that 
individuals are fit for duty, trustworthy, and reliable, and can safely 
and competently perform the duties and responsibilities making them 
subject to the FFD program. If not, the individuals would be assessed 
in accordance with the licensee's or other entity's procedures similar 
in manner to that required by subpart K of part 26, and the proposed 
PMRP would require corrective actions should a threshold be exceeded.
    If a licensee or other entity elects not to perform the analysis in 
proposed Sec.  26.603(c) to determine whether it satisfies the 
criterion in proposed Sec.  53.860(a)(2); performs the analysis and 
finds that the facility and its operation does not satisfy the 
criterion in proposed Sec.  26.603(c); or is a holder of an ML, the 
licensee or other entity could not implement the FFD program described 
in Sec.  26.604. Instead, the licensee or other entity would implement 
either the program described in proposed Sec.  26.605 or an FFD program 
that demonstrates compliance with all the requirements in current 
subparts A through I, N, and O of part 26.
    Proposed Sec.  26.605 would establish requirements in a graded 
manner similar to the regulatory framework established by the 
requirements in subparts A through I, N, O, and K of part 26. This 
existing graded approach consists of an FFD program for construction of 
a commercial nuclear plant and a more robust program that must be 
implemented before reactor operation. The former is the FFD program in 
proposed Sec.  26.605(a), and the latter is proposed Sec.  26.605(b). 
Like that for an FFD program under Sec.  26.604, the FFD program under 
Sec.  26.605 would include FFD program elements similar to those in 
subpart B of part 26, but the proposed requirements are less 
prescriptive, enabling more flexibility in program implementation like 
that offered in subpart K of part 26. For example, the requirements in 
subpart B of part 26 are explicit requirements for, in part, the 
collection and analysis of urine specimens. Subpart B of part 26 does 
not enable the use of oral fluid for drug testing or screening, except 
under very limited situations as described in subpart E of part 26, or 
the use of hair specimens, unlike proposed Sec.  26.605. Proposed Sec.  
26.605 would require drug and alcohol testing based on either the 
requirements in part 26 or the HHS Guidelines. The principal benefit of 
the proposed Sec.  26.605 FFD program is that it would provide a 
regulatory framework that is consistent with the radiological 
consequences for a facility that does not satisfy the criterion in 
proposed Sec.  53.860(a)(2) while affording flexibilities in the 
conduct of drug and alcohol testing.
    Proposed Sec.  26.605(a) would apply to licensees and other 
entities who perform the Sec.  26.603(c) analysis and satisfy the 
criterion in Sec.  53.860(a)(2) but decide not to implement the FFD 
program described in proposed Sec.  26.604, licensees and other 
entities who do not perform the Sec.  26.603(c) analysis, and licensees 
and other entities who perform the analysis but their analysis does not 
demonstrate that their facility and its operation satisfy the criterion 
in Sec.  53.860(a)(2). These entities must establish, implement, and 
maintain an FFD program under Sec.  26.605(a) either during 
construction activities as defined in Sec.  26.5, or during activities 
performed under an ML that allows the assembly, testing, or both, of a 
manufactured reactor. This FFD program implements all the FFD program 
requirements in Sec.  26.604 plus drug and alcohol testing.
    The timing element of the proposed applicability statement of Sec.  
26.605(a) is equivalent to that for an LWR licensee or other entity who 
is performing those same activities at a facility licensed under part 
50 or 52 and helps provide assurance that those individuals who 
assemble, test, or perform construction activities as defined in Sec.  
26.5 or direct these activities are fit for duty and trustworthy and 
reliable. This is important because assembly and testing a manufactured 
reactor and the construction and testing of SSCs required for facility 
operation require, in part, adherence to procedures, possible 
implementation of unique and precise assembly techniques, and quality 
assurance and controls. Additionally, SSCs within a manufactured 
reactor may not be accessible, testable, or available for quality 
assurance and verification after the reactor is assembled. This 
requirement is also proposed to address solo-assembly activities that 
may cause latent failures and passive SSCs located internal to a 
reactor (for example, a fusible link designed to melt at a particular 
temperature to trigger an actuation mechanism) that are relied upon for 
safe operation but cannot be inspected or tested for proper 
installation, configuration, or operation after installation. A Sec.  
26.605(a) FFD program for these types of activities is equivalent to 
the FFD program applicable to the assembly of the reactor vessel 
internals and testing of the SSCs internal to the reactor at an LWR 
licensed under part 50 or 52.

[[Page 86963]]

    Proposed Sec.  26.605(b) would apply to the same licensees and 
other entities as in proposed Sec.  26.605(a) but before the loading of 
fuel onsite into a reactor vessel; before receiving a manufactured 
reactor; or before individuals subject to part 26 operate, test, 
perform maintenance of, or direct the maintenance or surveillance of 
security-related equipment or equipment that a risk-informed evaluation 
process has shown to be significant to public health and safety. These 
entities must establish, implement, and maintain an FFD program that 
implements all the requirements in Sec.  26.605(a), except proposed 
Sec. Sec.  26.610, ``Sanctions''; 26.617, ``Recordkeeping and 
reporting''; and 26.619, ``Suitability and fitness determinations''; 
plus additional requirements due to the increased radiological 
consequences presented by a part 53 commercial nuclear plant as the 
licensee readies it for operation. These additional requirements 
include those in subparts C, D, H, and N of part 26, some of which 
would replace Sec. Sec.  26.610, 26.617, and 26.619.
    Proposed Sec.  26.605(b) would also enable the licensee or other 
entity to better integrate its facility into the LWR fleet and Category 
I fuel cycle facilities because subparts C, D, and H of part 26 would 
be required. These subparts would be required, in part, because it is 
expected that: (1) individuals will be able to work at any part 50, 52, 
or 53 commercial nuclear plant and will possess a nuclear safety 
culture and desirable qualifications, skills, expertise, or services; 
and (2) licensees and other entities of facilities licensed under parts 
50, 52, and 70 may venture to construct or operate a facility licensed 
under part 53. Therefore, the implementation of these subparts would 
help ensure that all individuals subject to part 26, except those 
individuals subject to an FFD program under Sec.  26.604, Sec.  
26.605(a), or subpart K of part 26, would be subject to FFD programs 
that provide reasonable assurance that the individuals are fit for 
duty, trustworthy, and reliable.
    Proposed Sec.  26.606, ``Written policy and procedures,'' would 
require licensees and other entities to implement and maintain an FFD 
policy and procedures for their FFD programs. This section would 
establish requirements equivalent to those in current Sec.  26.403, 
``Written policy and procedures,'' of subpart K. However, a principal 
difference is that proposed Sec.  26.606 is written to enable the use 
of urine, oral fluid, and hair for drug testing and screening.
    Proposed Sec.  26.606(a)(1) would require each licensee and other 
entity to provide a written FFD policy statement to individuals subject 
to the FFD program before the individuals are subjected to behavioral 
observation and any FFD program drug and alcohol test. This would be a 
protection measure afforded to individuals subject to the FFD program 
to help ensure that they know what is expected of them before being 
subject to the FFD program and potential consequences should they 
violate the FFD policy or procedures. This requirement would also 
contribute to safety and security because understanding FFD program 
responsibilities may enhance an individual's safety culture or the 
individual may self-select out of the licensee's or other entity's 
hiring process.
    Proposed Sec.  26.606(a)(2) would require that the FFD policy 
statement describe the performance objectives in Sec.  26.23, which are 
the same FFD program performance objectives required for facilities 
licensed under parts 50, 52, or 70. Having a standard performance 
outcome based on a licensee or other entity satisfying the Sec.  26.23 
performance objectives would enhance consistency in FFD program 
implementation across all entities subject to part 26. It would also 
generate confidence that individuals subject to part 26 will safely and 
competently perform their duties and responsibilities and use NRC-
licensed materials in a manner that will protect the public health and 
safety and common defense and security.
    Proposed Sec.  26.606(a)(3) would require that the FFD policy 
statement describe the minimum days off requirements in Sec.  
26.205(d)(3) or maximum average work hours requirements in Sec.  
26.205(d)(7).
    Proposed Sec.  26.606(a)(4) would require the FFD policy statement 
be written in sufficient detail to provide affected individuals with 
information on what is expected of them and what consequences may 
result from a lack of adherence to the policy, including those elements 
described in Sec.  26.603(b), part 26-required sanctions, and required 
medical/clinical treatment and follow-up testing for FFD policy 
violations. This requirement is equivalent to Sec.  26.403(a) of 
subpart K but includes an additional description of what the policy 
statement must include. For example, the policy would describe the NRC-
required sanctions to help deter substance abuse and required medical/
clinical treatment and follow-up testing for FFD policy violations. 
This provision would provide a protection measure by helping the 
individual get the assistance they need and help ensure that the 
individual refrains from substance abuse.
    Proposed Sec.  26.606(a)(5) would require that the FFD policy 
statement describes the individual's responsibilities to report for 
work in a physiological and psychological condition that enables the 
safe and competent performance of assigned duties and responsibilities 
and inform a licensee- or other entity-designated representative when 
the individual determines that this cannot be accomplished.
    Proposed Sec.  26.606(b) would require licensees and other entities 
implementing a FFD program in accordance with subpart M of part 26 to 
establish, implement, and maintain written procedures for their FFD 
programs. This requirement would be equivalent to that in Sec.  
26.403(b) of subpart K.
    Proposed Sec.  26.606(b)(1) would establish requirements for a 
subpart M of part 26 FFD program in which the licensee or other entity 
implements a drug and alcohol testing program. This provision would be 
equivalent to the requirements in current Sec.  26.403(b)(1) of subpart 
K, but Sec.  26.606(b)(1)(i) through (iv) proposes additional clarity 
and specificity that licensees and other entities must detail in their 
procedures to address new testing methods in subpart M of part 26 that 
are not permitted under the current part 26 framework. Clarity and 
specificity in procedural instructions would support consistent program 
implementation, which protects all individuals subject to the program.
    Proposed Sec.  26.606(b)(1)(iv) would require that if the licensee 
or other entity elects to use the HHS Guidelines for the conduct of 
drug testing, the FFD program procedures must include the name of the 
specific HHS Guideline and revision being implemented by the licensee 
or other entity and a description of the specific sections in the 
guideline that are being implemented, including specimen collections, 
drug testing, laboratory procedures, and evaluation of test results. 
This requirement would help ensure the following: the validity and 
accuracy of drug testing because the specimens would be subject to 
laboratory testing that has been certified by the HHS; protection of 
worker rights equivalent to the privacy, information, and due process 
protections afforded to Federal workers under the HHS Guidelines 
because the HHS Guidelines are used in the Federally mandated drug 
testing programs; consistency in program implementation because all 
individuals subject to the FFD program would be subject to the same 
collection,

[[Page 86964]]

testing, and evaluation processes; and FFD program effectiveness 
because the effectiveness of the HHS Guidelines have been verified by 
HHS's National Laboratory Certification Program (NLCP). Detailed 
procedures would enhance MRO and FFD program personnel reviews of 
individual test results because instructions would be provided for, in 
part, the evaluation of specific test results (e.g., positive, 
negative, biological markers), the conduct of additional testing for 
invalid or dilute specimens, and the assessment of subversion attempts 
(e.g., adulterated or substituted). This would benefit FFD program 
effectiveness and help prevent misunderstanding of program requirements 
and processes.
    Proposed Sec.  26.606(b)(2) would require licensees and other 
entities to include in their written procedures the immediate and 
follow-up actions that would be taken, and the procedures that would be 
used, in certain situations specified in proposed Sec.  26.606(b)(2)(i) 
through (vi). Proposed Sec.  26.606(b)(2) would be equivalent to the 
requirements in current Sec.  26.403(b)(2), which provides the same 
requirement under an FFD program for construction for part 50 or 52 
licensees and other entities. This would help ensure the effectiveness 
of the FFD program and its consistent implementation, because part 53 
licensed facilities would be implementing procedures to address the 
same requirements and with individuals who would understand what is 
expected of them no matter what part 53 facility they were assigned.
    The situation specified in proposed Sec.  26.606(b)(2)(i) would 
arise when individuals subject to the FFD program have been involved in 
the use, sale, or possession of illegal substances, illegal drugs, or 
illicit substances. This provision would be equivalent to current Sec.  
26.403(b)(2)(i), except that the phrase ``illegal drugs'' would be 
replaced with ``illegal substances, illegal drugs, or illicit 
substances.'' Illegal substances would include legal substances used in 
a manner inconsistent with Federal or State law.
    The situation specified in proposed Sec.  26.606(b)(2)(ii) would 
arise when individuals who are subject to the FFD program are impaired 
by any substance or the consumption of alcohol as determined by 
behavioral observation or a test that measures blood alcohol 
concentration, as defined in Sec.  26.5. Except for a few differences, 
this provision would be equivalent to current Sec.  26.403(b)(2)(ii) of 
subpart K. The NRC would not include the phrases ``to excess'' and 
``accurately'' in proposed Sec.  26.606(b)(2)(ii). Subpart M of part 26 
is a performance-based framework that focuses on impaired human 
performance, and for alcohol, impairment is determined by behavioral 
observation or by blood alcohol concentrations exceeding the limits in 
Sec.  26.103, ``Determining a confirmed positive test result for 
alcohol,'' using an evidentiary breath testing (EBT) device for alcohol 
(not whether an individual drank ``to excess''). If impairment is 
determined by an individual's behavior, it must be based on 
physiological indications of alcohol impairment. These indications are 
well established in medical, clinical, and law enforcement 
organizations, and could be used by the licensee or other entity 
through its procedures and training.\11\
---------------------------------------------------------------------------

    \11\ By ``well established'' the NRC means that there are 
Federal, State, and non-governmental organizations with reputable 
and scientifically based resources available for a licensee or other 
entity to use in its procedures or training to inform individuals of 
the physiological indications of alcohol impairment or intoxication.
---------------------------------------------------------------------------

    The NRC would include the phrase ``illegal substances, illegal 
drugs, and illicit substances'' in proposed Sec.  26.606(b)(2)(ii) 
based on operating experience and the terminology in current Sec.  
26.23(b). There are far more substances that may cause impairment than 
just drugs, drug metabolites, and alcohol. The phrase ``before or while 
constructing or directing construction of safety- or security-related 
SSCs'' in current Sec.  26.403(b)(2)(ii) would not be included in 
proposed Sec.  26.606(b)(2)(ii) because proposed Sec.  26.606 would 
apply during construction, operation, and decommissioning, if 
applicable. The NRC would include the term ``behavioral observation'' 
in proposed Sec.  26.606(b)(2)(ii) because impairment can be visibly or 
audibly observed in an individual, and individuals subject to subpart M 
of part 26 would be trained in behavioral observation under proposed 
Sec.  26.608.
    The situation specified in proposed Sec.  26.606(b)(2)(iii) would 
arise when individuals who are subject to an FFD program that includes 
drug and alcohol testing attempt to subvert the testing process by 
adulterating or diluting specimens (in vivo or in vitro), substituting 
specimens, or by any other means. Except for one difference, this 
provision would be equivalent to current Sec.  26.403(b)(2)(iii). The 
NRC would include the phrase ``if drug and alcohol testing is 
conducted'' to address the licensee or other entity who implements 
Sec.  26.604, which does not require drug and alcohol testing. The 
purpose underlying this requirement has increased in significance since 
issuance of the 2008 part 26 final rule because subversion attempts 
have accounted for about one-third of all FFD policy violations every 
year since 2016.
    The situation specified in proposed Sec.  26.606(b)(2)(iv) would 
arise when individuals, who are subject to an FFD program that includes 
drug and alcohol testing, refuse to provide a specimen for analysis or 
refuse to follow instructions provided by FFD program personnel. Except 
for two differences, this provision would be equivalent to current 
Sec.  26.403(b)(2)(iv). As with proposed Sec.  26.606(b)(2)(iii), the 
NRC would include the phrase, ``if drug or alcohol testing is 
conducted,'' to account for an FFD program implemented under Sec.  
26.604. The NRC would include the phrase ``or follow the instructions 
provided by FFD program personnel'' based on an existing requirement in 
Sec.  26.89(c) that the collector must inform the donor that if the 
donor refuses to cooperate in the specimen collection process, then 
such refusal will be considered a refusal to test and sanctions for 
subverting the testing process will be imposed.
    The situation specified in proposed Sec.  26.606(b)(2)(v) would 
arise when individuals who are subject to an FFD program had legal 
action taken relating to drug or alcohol use. This requirement would be 
equivalent to current Sec.  26.403(b)(2)(v).
    The situation specified in proposed Sec.  26.606(b)(2)(vi) would be 
when individuals subject to an FFD program demonstrated character or 
actions indicating that the individual cannot be trusted or relied upon 
to perform those duties and responsibilities or maintain access to NRC-
licensed facilities, SNM, or sensitive information. This includes 
character traits beyond those attributed to drug or alcohol use. This 
proposal would help ensure that the licensee or other entity will 
implement an FFD program designed to demonstrate compliance with the 
Sec.  26.23(c) performance objective that FFD programs must provide 
``reasonable measures for the early detection of individuals who are 
not fit to perform the duties that require them to be subject to the 
FFD program.'' An individual who is not trustworthy and reliable is not 
fit to perform or direct the performance of those duties and 
responsibilities or be afforded those types of access that make the 
individual subject to an FFD program.
    This proposed requirement also would help to align the subpart M of 
part 26 BOP with the BOP implemented under Sec.  73.56(f) and proposed 
Sec.  73.120 and the purpose of the IMP as described in Sec.  
73.55(b)(9) and proposed

[[Page 86965]]

Sec.  73.100(b)(9).\12\ The demonstrated character and actions of an 
individual can indicate whether the individual can be trusted and 
relied upon to safely and competently perform assigned duties and 
responsibilities or be afforded those types of access making the 
individual subject to the FFD program. This holds true for any 
demonstrated adverse character indication or action on- or offsite.
---------------------------------------------------------------------------

    \12\ The IMP must monitor the initial and continuing 
trustworthiness and reliability of individuals granted or retaining 
unescorted AA to a protected or vital area and implement defense-in-
depth methodologies to minimize the potential for an insider to 
adversely affect, either directly or indirectly, the licensee's 
capability to protect against radiological sabotage.
---------------------------------------------------------------------------

    The phrase ``character or actions'' would be used in proposed Sec.  
26.606(b)(2)(vi) to focus on observed examples that indicate an 
individual subject to subpart M of part 26 may not be fit for duty or 
trustworthy and reliable. Character traits include but are not limited 
to personality, temperament, honesty, carelessness, apathy, psychosis, 
and commitment to safety culture. Assessment of an individual's 
character should consider the potential for changes in these traits 
when compared to a previous baseline. Actions would include a physical 
or verbal demonstration of a character trait that could call into 
question an individual's fitness, trustworthiness, or reliability. For 
example, the individual does something physically, verbally, or in 
writing (e.g., falsifying records, driving while impaired, or harming 
or threatening to harm oneself, others, or property) that compels 
another individual to conclude that the observed individual cannot be 
trusted or relied upon. Unlike the background investigation and reviews 
of ``character and reputation'' in Sec.  73.56(d)(6) and (k)(1)(v) and 
proposed Sec.  73.120, which are principally retrospective reviews of 
an individual and may be based on third-party information (i.e., 
information from individuals not subject to NRC requirements), the 
``character or action'' focus of proposed Sec.  26.606(b)(2)(vi) would 
be a present observation of an individual subject to the FFD program 
and performed by an individual who is also subject to the FFD program. 
Whether the information would be received from an individual subject to 
the FFD program or someone who is not subject to the FFD program, the 
licensee or other entity would need to review this information (i.e., 
determine if the information and its source are credible) to determine 
whether the individual should maintain authorization.
    Proposed Sec.  26.606(b)(3) would require licensees and other 
entities to address in their procedures the process, including the 
duties and responsibilities of FFD program personnel, to be followed if 
an individual's behavior or condition raises an FFD concern. This 
provision would also require a process to be conducted when credible 
information is received by the licensee or other entity that the 
individual is not fit for duty, trustworthy, and reliable.
    With a few exceptions, proposed Sec.  26.606(b)(3) would be 
equivalent to current Sec.  26.403(b)(3). Instead of the phrase ``while 
constructing or directing the construction of safety- or security-
related SSCs'' in current Sec.  26.403(b)(3), the NRC would use ``on 
the NRC-licensed facility'' in proposed Sec.  26.606(b)(3) because this 
provision would apply during commercial nuclear plant construction, 
operation, and decommissioning, if applicable, in addition to holders 
of an ML as described in Sec.  26.3(f). The requirement that the roles 
and responsibilities of FFD program personnel be described was 
developed from current Sec. Sec.  26.4(g) and 26.31(b) and operating 
experience, which has demonstrated that clear job descriptions help 
ensure that individuals know who is designated by the licensee or other 
entity to make decisions regarding FFD program implementation and who 
can be approached when physiological or psychological help is needed. 
This is principally a protection consideration afforded to individuals 
subject to the FFD program.
    The proposed requirement would also include two conditions not 
found in current Sec.  26.403(b) that would clarify the initiation of 
the fitness determination process should an individual's behavior or 
condition raise an FFD concern. The phrase, ``impairment from any cause 
that in any way could adversely affect the individual's ability to 
safely and competently perform the individual's duties,'' would reflect 
the Sec.  26.23(b) performance objective. The condition, ``the receipt 
of credible information indicating that the individual cannot be 
trusted or relied on to perform those duties and responsibilities 
making the individual subject to this part,'' would reflect the Sec.  
26.23(a) performance objective. In either case, as required by Sec.  
26.23(c), the FFD program must provide reasonable measures for the 
early detection of individuals who are not fit to perform the duties 
that require them to be subject to the FFD program.
    Proposed Sec.  26.606(b)(4) would require licensees and other 
entities to have written procedures that address the operation and 
oversight of an onsite or offsite collection facility. This requirement 
would be equivalent to current Sec. Sec.  26.403(b) and 26.405(e) and 
is developed from Sec.  26.41(b), which states that each licensee and 
other entity who is subject to subpart B of part 26, shall ensure that 
the entire FFD program is audited, which is part of a licensee's or 
other entity's oversight of the facility, and Sec.  26.87(a), which 
states that each FFD program must have one or more designated 
collection sites that have all necessary personnel, materials, 
equipment, facilities, and supervision to collect specimens for drug 
testing and to perform alcohol testing. Having procedures for the 
operation and oversight of the onsite or offsite collection facility 
would enhance consistency in program implementation, protect 
individuals subject to testing, and account for the flexibilities 
afforded in the types of biological specimens than may be collected 
under an FFD program subject to subpart M of part 26. Section 
26.606(b)(4), when used with the PMRP described in Sec.  26.603(d) and 
the proposed audit requirement in Sec.  26.605(a), would help maintain 
FFD program effectiveness and prevent subversion attempts at facilities 
that may not be under the direct day-to-day oversight of FFD program 
personnel.
    Proposed Sec.  26.606(b)(5) would require licensees and other 
entities to have written procedures that address the fatigue management 
requirements in Sec.  26.202(b), ``Procedures,'' and either Sec.  
26.205(d)(3) or (d)(7).
    Proposed Sec.  26.606(b)(6) would require licensees and other 
entities to have written procedures that provide measures to prevent 
subversion of drug and alcohol tests conducted onsite and offsite. This 
proposal was developed from Sec.  26.27(c)(1).
    Proposed Sec.  26.607, ``Drug and alcohol testing,'' would 
establish drug and alcohol testing requirements for licensees and other 
entities implementing proposed Sec.  26.604, at their discretion, and 
licensees and other entities implementing proposed Sec.  26.605. Except 
for a few differences, proposed Sec.  26.607 would be equivalent to 
current Sec.  26.405, which requires licensees and other entities 
implementing an FFD program under subpart K of part 26 to have a drug 
and alcohol testing program that demonstrates compliance with the 
requirements in Sec.  26.405(b) through (g). The differences are 
commensurate with the risk consequences presented by a part 53-licensed 
facility as compared to a part 50 or 52 nuclear power plant. These 
proposed requirements would improve flexibility in the conduct of

[[Page 86966]]

drug and alcohol testing while maintaining protections afforded to 
individuals subject to the FFD program.
    Proposed Sec.  26.607(a) would require licensees and other entities 
to obtain a split specimen for all drug tests using oral fluid or urine 
for all test conditions in Sec.  26.607(b), (h) and (j). Neither 
current subpart K nor current subparts B or E of part 26 require a 
split specimen. However, the majority of the LWR fleet uses split 
specimens for drug testing and commercially available drug screening 
products use a split specimen technique. Since publication of the 2008 
part 26 final rule, the HHS has issued guidelines for urine and oral 
fluid that require split specimens, and the draft proposed HHS 
Guidelines for hair requires split specimens, as well.
    The required use of a split specimen process would protect the 
individual because, upon a donor-alleged discrepant or questionable 
test result, the donor may provide permission to test the split 
specimen (specimen B) in an effort to refute the laboratory test 
results for specimen A. The requirement also would enable the MRO to 
direct laboratory testing of specimen B if specimen A were invalid; 
though the NRC expects specimens becoming invalid at the laboratory to 
be a rare occurrence as testing would be conducted in HHS-certified 
laboratories with trained collectors. In the event that a specimen is 
determined to be invalid, then the occurrence would likely warrant 
further investigation by the MRO and laboratory to identify the cause. 
This protocol would be equivalent to the special analysis testing in 
current Sec.  26.163(a)(2) for dilute specimens in that additional 
laboratory analysis is performed because of a questionable test result.
    If a split specimen is tested by an HHS-certified laboratory, then 
the test result from specimen B must be used as part of the 
determination for an FFD policy violation as required by Sec.  
26.185(n), ``Evaluating results from a second laboratory.'' However, 
this is not to say that the test results from specimen A should be 
discarded. Since the HHS-certified laboratory should report all test 
results from all specimens tested to the MRO, like the information 
described in Sec.  26.169, ``Reporting results,'' test result 
differences between specimens A and B can be used to inform the MRO as 
to what should be reported to the licensee or other entity to either 
facilitate medical or clinical assistance for the individual, inform an 
FFD-policy violation determination, or both.
    The proposed Sec.  26.607(a) requirement would also state that if 
the licensee or other entity elects to use a POCTA device for screening 
during random testing or portal area monitoring (e.g., pre-access 
screening), a split specimen would not need to be taken. The reason for 
this exception would be that the requirements in Sec.  26.607(h)(4) 
establish the process to be implemented when a screening test indicates 
a presumptive positive, adulterant, or a discrepant biological marker, 
if applicable. This process includes collecting and testing a specimen 
for analysis at an HHS-certified laboratory.
    Proposed Sec.  26.607(b) would require the licensee or other entity 
to subject individuals identified in Sec.  26.202 to drug and alcohol 
testing under the five conditions listed in Sec.  26.607(b)(1) through 
(5). Proposed Sec.  26.607(b) would be equivalent to current Sec.  
26.405(c).
    Proposed Sec.  26.607(b)(1) would require pre-access testing 
similar to current Sec.  26.405(c)(1), which requires testing before 
assignment to construct or direct the construction of safety- or 
security-related SSCs. Unlike current Sec.  26.405(c)(1), the proposed 
requirement would not include the phrase, ``construct or direct the 
construction of safety- or security-related SSCs,'' because, for 
licensees or other entities under part 53, the pre-access test 
condition applies to construction, operation, and decommissioning, if 
applicable, to help inform a licensee's or other entity's authorization 
determination. The proposal also would use ``pre-access'' instead of 
``pre-assignment,'' which is used in current Sec.  26.405(c)(1).
    A pre-access test would require the collection of an oral fluid or 
a urine specimen no more than 14 days before the individual is granted 
unescorted access. Although this change has roots in the 2008 part 26 
final rule, which reduced the period within which pre-access testing 
must be performed from 60 days to 30 days or less, the 14-day proposal 
is based on three lessons learned from operating experience.
    First, the 14-day period would be a large enough window of time to 
collect the specimen and evaluate test results because licensees or 
other entities typically receive laboratory test results within 5 
business days of laboratory receipt of the biological specimen. At the 
same time, the 14-day period would be small enough to help ensure that 
the test results are representative of the individual's forensic 
toxicology before being granted authorization.
    Second, the 14-day window would enable the licensee or other entity 
to conduct an unannounced pre-access drug and alcohol screening using a 
hair specimen or a POCTA. This would help prevent an individual from 
attempting to subvert the drug and alcohol test by temporarily 
abstaining from drug or alcohol abuse or adulterating or substituting 
their specimen to obtain a non-positive test result.
    Third, the NRC does not expect licensees and other entities 
licensed under part 53 to have the large and periodic influxes of 
individuals (either licensee employees or C/Vs) that LWRs have to 
support facility operation, maintenance, engineering design changes, or 
nuclear refueling. Therefore, these licensees or other entities would 
not be periodically challenged to in-take a large workforce within the 
proposed 14-day pre-access testing window.
    Proposed Sec.  26.607(b)(2) would require the licensee or other 
entity to conduct random drug and alcohol testing of all individuals 
subject to the FFD program. With one exception, this proposed 
requirement would be equivalent to current Sec.  26.405(b). Section 
26.405(b) gives licensees and other entities that implement an FFD 
program subject to subpart K of part 26 the option to impose random 
drug and alcohol testing. Proposed Sec.  26.607(b)(2) would not offer 
that option because subpart M of part 26, unlike subpart K, would not 
allow a licensee or other entity to implement a fitness monitoring 
program under current Sec.  26.406 instead of a random testing program. 
The principal reasons for not allowing this flexibility would be that 
no licensee or other entity has ever implemented a fitness monitoring 
program (i.e., there is no operating or regulatory experience on which 
to judge the effectiveness of a fitness monitoring program) and the 
proposed subpart M framework already uses behavioral observation to 
help ensure FFD program effectiveness. Supplementing the proposed Sec.  
26.609 BOP with an additional observation technique (i.e., the fitness 
monitoring program) would not result in a level of deterrence or 
detection equivalent to that which would be obtained through behavioral 
observation and random drug and alcohol testing.
    Proposed Sec.  26.607(b)(2)(i) through (v) would provide specific 
requirements for the conduct of a random testing program. These 
paragraphs would be equivalent to Sec.  26.405(b)(1) through (4), 
although with a few differences. The similar provisions would be 
proposed in Sec.  26.607(b)(2)(i), (b)(2)(iii), and (b)(2)(iv).
    The differing provisions would include proposed Sec.  
26.607(b)(2)(ii), which would refer to an ``FFD program procedure'' 
instead of the reference to an ``FFD program policy'' in Sec.  
26.405(b)(2) because procedures

[[Page 86967]]

contain the instructions that implement FFD program requirements, but 
the FFD policy need not contain specific instructions. Section 
26.607(b)(2)(ii) would also require individuals who are selected for 
random testing to report to the onsite collection site, as opposed to 
the collection site in Sec.  26.405(b)(2) because alcohol metabolism 
necessitates a relatively timely alcohol test. This change is also 
proposed because the NRC expects that part 53 licensees and other 
entities may use a combination of onsite (for random, for-cause, and 
post-event testing) and offsite (for pre-access, post-event, and 
follow-up testing) collection facilities for drug and alcohol testing 
and may have to afford reasonable accommodation to certain individuals, 
which would add complexity in the licensee's or other entity's 
procedurally determined time period in which an individual must report 
to the collection facility.
    Another difference from Sec.  26.405(b) would be proposed Sec.  
26.607(b)(2)(v), which would establish the random testing rate for the 
population of individuals subject to testing. Subpart K of part 26 does 
not establish a random testing rate. The proposed requirement would be 
equivalent to current Sec.  26.31(d)(2)(vii), which requires that the 
sampling process used to select individuals for random testing provides 
that the number of random tests performed annually is equal to at least 
50 percent of the population that is subject to the FFD program. The 
NRC would revise that slightly for proposed Sec.  26.607(b)(2)(v) to 
require a 50 percent random testing rate for the licensee employee 
population and a 50 percent random testing rate for the C/V population. 
The NRC proposes this change for two reasons.
    First, although operating experience has demonstrated that Sec.  
26.31(d)(2)(vii) helps provide reasonable assurance that individuals 
are fit for duty and trustworthy and reliable through the detection and 
deterrence of substance abuse, this same operating experience 
demonstrates that, on many occasions, the C/V population has been 
tested at a rate lower than 50 percent, even though this population 
results in the majority of all FFD policy violations. This bias occurs 
because C/Vs are available for testing only during short periods of 
time or periodically throughout the year, whereas licensee employees 
are essentially always available for a test.
    A second reason why the NRC is proposing a different 50 percent 
random testing protocol than in the current part 26 requirements is 
that the flexibilities afforded to part 53 licensees or other entities 
in subpart M of part 26 are not afforded to licensees or other entities 
that must implement an FFD program under subparts A through I, N, and O 
of part 26. These flexibilities include enabling the use of a POCTA 
device to screen individuals during the random testing process and the 
use of offsite collection facilities for pre-access testing. The 
potential reduction in FFD program effectiveness caused by licensee or 
other entity implementation of these options would be offset by subpart 
M requirements that mitigate possible challenges to the FFD program, 
such as the 50 percent random testing rate for the licensee employee 
population and 50 percent random testing rate for the C/V population.
    Proposed Sec.  26.607(b)(3) would require for-cause testing 
equivalent to that used in current FFD programs implementing Sec.  
26.405(c)(2). The NRC would require for-cause testing, like random 
testing, to be conducted onsite to ensure that the test is conducted as 
soon as reasonably practicable. This is an important consideration when 
for-cause testing for alcohol or using oral fluid for drug screening or 
testing because human metabolism continually lowers the concentrations 
of the drugs, drug metabolites, and alcohol perhaps to concentrations 
lower than the initial or confirmatory testing cutoffs. Additionally, 
for facilities that are sited in geographically remote locations, an 
offsite collection facility might be too far away or not readily 
accessible.
    Proposed Sec.  26.607(b)(4) would require post-event testing in a 
manner equivalent to current Sec.  26.405(c)(3) with a few adjustments. 
For part 53 licensees or other entities, the NRC proposes post-event 
testing under two conditions: events involving human errors that may 
have caused or contributed to the events (proposed Sec.  
26.607(b)(4)(i)), and events not involving human error that result in 
adverse health consequences or damage to any safety- or security-
related SSC (proposed Sec.  26.607(b)(4)(ii)). The word ``significant'' 
would not be used in Sec.  26.607(b)(4)(ii)(A) to describe the 
``illness or personal injury'' as used in Sec.  26.405(c)(3)(i) because 
Sec.  26.607(b)(4)(ii)(A) would describe which illnesses or injuries 
are covered. Proposed Sec.  26.607(b)(4)(ii)(B), unlike Sec.  
26.405(c)(3)(ii), would not use the word ``significant'' to describe 
the damage to safety- or security-related SSCs because any damage to 
safety- or security-related SSCs would require testing within four 
hours of the event unless immediate medical intervention precludes the 
conduct of the test on the individual(s) who caused or contributed to 
the event. Proposed Sec.  26.607(b)(4)(ii)(B) also would not use the 
word ``construction'' as in Sec.  26.405(c)(3)(ii) because Sec.  
26.607(b)(4) would apply to construction, operation, and 
decommissioning, if applicable.
    Proposed Sec.  26.607(b)(4)(i) would require the licensee or other 
entity to define in its procedures the terms ``human error'' and 
``event.'' These terms may take on various meanings and they are not 
defined in the current or proposed rule, so the licensee or other 
entity would be required to describe or define these terms to help 
ensure consistent implementation of subpart M of part 26 and that the 
post-event test condition would be consistently applied to all 
individuals subject to the FFD program. The Sec.  26.405(c)(3)(i) 
requirement that ``the event is recordable under the Department of 
Labor standards contained in 29 CFR 1904.7, and subsequent amendments 
thereto,'' would not be carried over to proposed Sec.  26.607(b)(4). 
Instead, the NRC proposes to prescribe the post-event test conditions 
in Sec.  26.607(b)(4), in part so they would not change unless the NRC 
amends the requirement.
    Proposed Sec.  26.607(b)(5) would require follow-up testing. This 
requirement would be equivalent to current Sec.  26.405(c)(4), although 
the proposed Sec.  26.607(b)(5) would further describe follow-up 
testing. The NRC proposes to describe follow-up testing as part of a 
series of tests for drugs, alcohol, or both, which are performed after 
an individual subject to part 26 has violated the FFD policy on 
substance use or abuse, or the sale, use, or possession of illegal 
drugs. Follow-up testing would be used to verify an individual's 
continued abstinence from substance abuse. The NRC would not include a 
reference to a follow-up plan as in Sec.  26.405(c)(4) because the 
intent of a follow-up plan is to conduct a series of drug tests, 
alcohol tests, or both, to verify continuing abstinence from substance 
abuse. Nevertheless, individuals who violate an FFD policy on substance 
use or abuse, or the sale, use, or possession of illegal drugs, should 
have a follow-up plan that includes a definition of ``abstinence'' from 
the medical professional prescribing the plan.
    Proposed Sec.  26.607(c) would provide additional testing 
requirements. This proposed requirement would be equivalent to Sec.  
26.405(d) and would require implementation of select requirements from 
current subpart E of part 26. The proposed requirements would govern 
directly observed collections, shy bladder situations, special analysis 
testing, and alcohol testing. These requirements would be necessary to 
maintain FFD program

[[Page 86968]]

effectiveness equivalent to that currently implemented by the LWR 
fleet.
    Proposed Sec.  26.607(c)(1) would require validity testing and 
establish the minimum panel of drugs and drug metabolites to be tested. 
This panel would be the same as those in Sec. Sec.  26.31(d)(1) and 
26.405(d) because, based on operating experience from LWR FFD program 
implementation, this panel has been determined to contribute to a 
licensee or other entity satisfying the FFD performance objectives in 
Sec.  26.23(a) through (d).
    Proposed Sec.  26.607(c)(1) would differ from Sec.  26.405(d) 
because it would require testing of oral fluid and urine specimens for 
validity, including at least one biological marker (developed from an 
HHS Guidelines provision) and one adulterant (equivalent to current 
validity testing for urine specimens in part 26). Section 26.405(d) 
requires that urine specimens collected for drug testing be subject to 
validity testing. The addition of oral fluid validity testing is 
important because, just as there are publicly available kits to subvert 
a urine drug test, kits that may be used to subvert a drug test that 
uses oral fluid as a biological specimen are also readily available.
    Proposed Sec.  26.607(c)(2) would include requirements that already 
exist in the part 26 framework that provide protections for individuals 
subject to the FFD program and contribute to testing effectiveness when 
collecting and assessing a urine specimen. Specifically, current Sec.  
26.115, ``Collecting a urine specimen under direct observation,'' 
describes the exclusive grounds for performing a directly observed 
collection and the process to be followed to protect the privacy of the 
individual. Section 26.119, ``Determining `shy' bladder,'' establishes 
the process to be followed when a donor is not able to produce a 
sufficient amount of urine for testing, and Sec.  26.163(a)(2) requires 
special analysis testing when a specimen is dilute to help prevent a 
subversion attempt.
    Proposed Sec.  26.607(c)(3) would require implementation of all the 
current alcohol testing requirements in Sec.  26.91, ``Acceptable 
devices for conducting initial and confirmatory tests for alcohol and 
methods of use,'' through Sec.  26.103, ``Determining a confirmed 
positive test result for alcohol.'' Using the same alcohol testing 
framework for parts 50, 52, 70, and 53 licensees and other entities 
would provide for regulatory consistency, protections for individuals 
subject to the FFD program (e.g., the quality controls and verification 
applied to the EBT device), and FFD program effectiveness (e.g., 
accuracy of test results). For alcohol testing, unlike drug testing, 
there is a preponderance of evidence that correlates blood alcohol 
concentrations to impairment and intoxication. Furthermore, FFD 
performance data has demonstrated that the time-dependent alcohol 
cutoffs in Sec.  26.103 have increased the detection of individuals who 
are under the influence of alcohol. For these reasons, the current 
alcohol requirements in part 26 are proposed for FFD programs under 
subpart M.
    Proposed Sec.  26.607(c)(4) would establish additional testing 
requirements. This proposal would be equivalent to current Sec.  
26.405(f) for facilities licensed under part 53 for the conduct of drug 
testing. Unlike Sec.  26.405(f), proposed Sec.  26.607(c)(4) would not 
reference validity screening and initial drug and validity tests at 
licensee testing facilities as this would be required in proposed Sec.  
26.607(c)(1). Another minor difference between Sec.  26.405(f) and 
proposed Sec.  26.607(c)(4) would reflect the requirement in subpart M 
of part 26 to use an HHS-certified laboratory for all biological 
specimens collected and not just for urine specimens.
    Consistent with Sec.  26.405(f), proposed Sec.  26.607(c)(4) would 
require the use of an HHS-certified laboratory for all test conditions 
listed in Sec.  26.607(b), MRO-directed tests, and the testing of a 
split specimen. Further, HHS-certified laboratory test results using 
urine or oral fluid would be required for the issuance of an FFD policy 
violation and part 26-required sanction.
    All drug testing would need to be performed at an HHS-certified 
laboratory to help ensure FFD program effectiveness and to protect the 
donor from a false positive test result and an unwarranted FFD policy 
violation. The donor would be protected because laboratory procedures 
for specimen accessioning, testing, custody and control, and evaluation 
of test results and the training and qualification of laboratory 
personnel are evaluated by HHS as part of the NLCP. This provides 
assurance that the drug testing results are accurate and attributed to 
the donor. Urine, oral fluid, and hair specimens may also be screened 
and tested for drugs and alcohol as described in Sec.  26.607. Drug and 
alcohol screening results obtained from urine and oral fluid specimens 
collected and analyzed using a POCTA device and screening results 
obtained from a hair specimen or a portal monitor may only be used as 
potentially disqualifying information for a licensee's or other 
entity's authorization determination (i.e., used to assess the fitness, 
trustworthiness, and reliability of the individual). These screening 
results may not be used for the administration of an FFD policy 
violation and sanction, except as proposed Sec. Sec.  26.607(i)(3) and 
26.610 for subversions, as defined in Sec.  26.5, of the drug and 
alcohol screening process.
    There are three phrases or requirements in Sec.  26.405(f) that the 
NRC does not propose to use in Sec.  26.607(c)(4). The first is the 
phrase, ``consistent with its standards and procedures for 
certification,'' regarding the operation of an HHS-certified 
laboratory, because the laboratory would not be HHS-certified if it 
were not following ``its standards and procedures for certification.'' 
The second is the requirement that urine specimens that yield positive, 
adulterated, substituted, or invalid initial validity or drug test 
results must be subject to confirmatory testing by the HHS-certified 
laboratory, except for invalid specimens that cannot be tested. This 
requirement would not be used because, under subpart M of part 26, 
licensees or other entities would be required to use an HHS-certified 
laboratory. For a laboratory to be HHS-certified, it must follow the 
HHS Guidelines and include procedures that describe when a specimen 
cannot be tested. Lastly, the Sec.  26.405(f) requirement that other 
specimens that yield positive initial drug test results must be subject 
to confirmatory testing by a laboratory that demonstrates compliance 
with stringent quality control requirements that are comparable to 
those required for certification by the HHS, would not be used because 
subpart M of part 26 would require the use of an HHS-certified 
laboratory.
    Proposed Sec.  26.607(c)(4) would require the licensee or other 
entity to contract with a primary and backup HHS-certified laboratory. 
This provision would help ensure that specimens are processed and 
tested to maintain FFD program effectiveness should the primary 
laboratory be unable to perform specimen testing. This would help 
maintain protections afforded to individuals subject to the FFD program 
(e.g., should the donor or MRO request testing of the split specimen, a 
different laboratory could be used). This requirement also would state 
that the primary and backup laboratories must have a different 
certifying scientist. Having a back-up HHS-certified laboratory and a 
different certifying scientist would benefit the program and donor 
because the drug testing instruments, technicians, and certifying 
scientist would be independent of the

[[Page 86969]]

primary laboratory testing and review process. The back-up HHS-
certified laboratory may be of the same corporate entity as the primary 
laboratory.
    Proposed Sec.  26.607(c)(4) would also state that the laboratory 
would be subject to inspection or audit by the licensee or other entity 
and that records and documents must be provided and/or able to be 
photocopied and removed from the premises to support the inspection or 
audit. This requirement would be equivalent to current Sec.  26.41(d) 
except that laboratories would not be able to limit the use and 
dissemination of documents copied or taken from the laboratory by a 
licensee or other entity. This is necessary to ensure the continuing 
effectiveness of FFD programs, because NLCP findings and audit results 
could adversely impact FFD program effectiveness. Pertinent information 
includes and should not be limited to NLCP-identified weaknesses (e.g., 
custody and control, accessioning, instrumentation, procedures, 
training, supervision, review of test results, and resolution of 
previously identified corrective actions) that may impact the 
effectiveness of FFD programs.
    Proposed Sec.  26.607(d) would help protect the donor from mistakes 
made during the drug and alcohol testing processes and help ensure FFD 
program effectiveness. The rule would require the licensee or other 
entity to protect the individual's privacy and the integrity of the 
specimen and to implement quality controls to ensure that test results 
are valid and attributable to the correct individual. This requirement 
would be equivalent to the first sentence of current Sec.  26.405(e), 
except that the word ``stringent'' was removed from the phrase 
``stringent quality controls,'' because the word ``stringent'' is not 
defined.
    Proposed Sec.  26.607(e) would describe the requirements for 
licensees and other entities that use offsite collection facilities. 
Consistent with current Sec.  26.405(e), a licensee or other entity 
would be able to conduct specimen collections and alcohol testing at a 
local hospital or other facility. Unlike Sec.  26.405(e), proposed 
Sec.  26.607(e) would not restrict licensees and other entities to use 
hospitals and other facilities that meet the requirements in 49 CFR 
part 40, ``Procedures for Transportation Workplace Drug and Alcohol 
Testing Programs,'' because subpart M of part 26 is intended to provide 
flexibilities beyond those in the current part 26 framework. Licensees 
and other entities may use these Department of Transportation 
requirements to inform their procedures under Sec.  26.606(b)(1) as 
long as the procedures do not conflict with the requirements in part 26 
or the HHS Guidelines.
    Proposed Sec.  26.607(e) would also require licensees and other 
entities to audit offsite collection facilities before their use and 
biennially to confirm that the facility procedures are comparable to 
those described in subpart E of part 26 or the HHS Guidelines for urine 
and oral fluid. This proposed requirement is based on current Sec.  
26.41(a) and (b). The Sec.  26.607(e) audit requirement would be a 
program effectiveness consideration because offsite collection 
facilities may not require vigilance of their collectors (e.g., 
identification of subversion attempts), diligence in the protection of 
worker rights (e.g., privacy and specimen custody and control), or 
procedural compliance.
    The offsite facility used by a licensee or other entity under 
proposed Sec.  26.607(e) would have to be licensed to conduct specimen 
collections and perform alcohol testing, and be audited, by the State 
or a State-designated entity. This requirement would help provide 
assurance of adequate collection facility performance and may help 
reduce the burden on the licensee or other entity and the collection 
facility. Crediting a State audit (or State licensure, oversight, or 
regulation) is established in Sec. Sec.  26.4(i)(4) and (j), 
26.91(e)(5), 26.153(f)(1), and 26.183(a).
    Proposed Sec.  26.607(f) would provide the requirements for initial 
drug testing. This provision would be equivalent to Sec.  26.405(f) 
except to account for the alternative biological specimens that may be 
tested under subpart M of part 26. For the testing of all biological 
specimens, the licensee or other entity under part 53 would be required 
to use a device that employs an immunoassay screening technique, or an 
alternative technology that the licensee or other entity has 
incorporated into its FFD program through the Sec.  26.603(e) change 
control process, that demonstrates compliance with the requirements of 
the U.S. Food and Drug Administration (FDA) for commercial 
distribution. Examples of alternative technologies include liquid or 
gas chromatography and mass spectrometry. Licensees and other entities 
would use the Sec.  26.603(e) change control process to evaluate and 
document a change to their collection and analysis procedures to enable 
the use of a better or perhaps more cost-effective collection and/or 
testing technology. Another difference from Sec.  26.405(f) would be 
changing the word ``urine'' in Sec.  26.405(f) to ``biological 
specimens'' in Sec.  26.607(f). Lastly, proposed Sec.  26.607(f) would 
include the phrase ``discrepant biological marker'' as a drug screening 
result that must be analyzed by an HHS-certified laboratory and 
evaluated by the MRO to help inform the MRO's determination of a 
subversion attempt.
    Proposed Sec.  26.607(g) would enable a part 53 licensee to use 
oral fluid as a biological specimen for testing. This requirement would 
be equivalent to Sec.  26.31(d)(5), which enables the MRO to conduct 
drug and alcohol testing using alternative methods, and Sec.  26.405, 
which does not preclude the use of oral fluid specimens for FFD 
programs that implement subpart K of part 26 requirements. In order to 
provide assurance that drug testing is effective and protects the 
worker, Sec.  26.607(g) would require that the licensee's or other 
entity's procedures incorporate the HHS Guidelines or the requirements 
in part 26 for the conduct of urine or oral fluid testing.
    The proposed Sec.  26.607(g) requires that the oral fluid 
collection device must have received premarket approval from the FDA 
and must not expire before laboratory testing. Also, the drugs, drug 
metabolites, initial and confirmatory testing cutoffs, and biological 
markers, if applicable, must be those established by HHS for oral fluid 
drug testing and the alcohol cutoffs in part 26. If they are not 
established by HHS or the NRC for the paneled drugs and drug 
metabolites, then they would be determined and documented by a forensic 
toxicologist review. This forensic toxicologist review would help 
ensure that the device accurately tests for the drug, drug metabolite, 
biological markers, adulterants, and/or alcohol and that the results 
from the device are comparable to those established in the HHS 
Guidelines for oral fluid testing.
    Proposed Sec.  26.607(h)(1) and (2) would enable the use of a POCTA 
device during the random and pre-access testing processes. These 
requirements are adopted from Sec.  26.97, ``Collecting oral fluid 
specimens for alcohol and drug testing,'' and Sec.  26.405(f), which 
does not preclude the use of oral fluid testing. To use a POCTA device 
for urine, oral fluid, or other biological indicators (breath, sweat, 
etc.), a forensic toxicology review would be required to ensure that 
the device is forensically effective. If the POCTA device is 
forensically effective, then the donor would be reasonably protected 
from a false positive test result, the licensee or other entity would 
be reasonably protected from false negative test results, and the FFD 
program would remain effective. For a POCTA device to be forensically 
effective, the forensic toxicologist would need to document an 
evaluation that the performance of the

[[Page 86970]]

POCTA device must be comparable to the requirements in Sec.  26.161(b) 
for a urine specimen or the procedures in the HHS Guidelines for urine 
or oral fluid, as implemented by the licensee or other entity through 
its procedures.
    The use of POCTA for oral fluid and urine specimens for the pre-
access and random testing processes would be acceptable because 
individuals in the pre-access process would be subject to an oral fluid 
or urine specimen collection and possible drug screening using a hair 
specimen, which are both required to be sent to an HHS-certified 
laboratory. For random testing, the individual would have also been 
granted authorization under the AA and FFD requirements and have been 
subject to behavioral observation and physical protection screening 
(e.g., verification of identify, and screening for explosives and 
contraband).
    Proposed Sec.  26.607(h)(3) would require that procedures be 
developed that ensure the effectiveness of the POCTA collection 
process, assessment of the screening results, and prevention of 
subversion attempts. This requirement would be equivalent to current 
Sec.  26.403(b)(1) and would help ensure protections afforded to 
individuals subject to the FFD program and program effectiveness. The 
subpart M of part 26 framework enables the use of POCTA for random 
screening of individuals for any part 53 facility, so the licensee or 
other entity should exercise due diligence and implement risk 
management strategies to ensure the efficacy of random screening and 
its contribution to an effective FFD program.
    Proposed Sec.  26.607(h)(4) would provide that an individual donor 
who screens positive (or whose specimen is invalid or indicates a 
discrepant biological marker or adulterant) is removed from all duties 
and responsibilities making the donor subject to subpart M of part 26. 
Under proposed Sec.  26.607(h)(4)(i), the donor then would be 
immediately subject to a drug and alcohol test that provides quantified 
confirmatory test results from which an FFD policy violation may be 
issued. Similar to other requirements for specimen collections, except 
for biological specimens analyzed by a passive detection system, the 
licensee or other entity would be required to implement procedures that 
ensure that all specimens collected are uniquely assigned to the donor 
(i.e., procedures that provide for custody and control of the 
specimen). If the individual shows signs of impairment during the POCTA 
process, proposed Sec.  26.607(h)(4)(ii) would require the temporary 
removal of the individual's authorization until the MRO reviews the 
laboratory test result(s), and interviews the individual, and a 
determination of fitness finds that authorization may be restored. 
Section 26.607(h)(4) is equivalent to Sec.  26.77(b) and was informed 
by the requirements in Sec. Sec.  26.419, 26.75(c) and (d), and 
26.185(c).
    Proposed Sec.  26.607(i) would enable the collection of hair 
specimens for drug testing to supplement pre-access testing that uses 
urine or oral fluid specimens. Hair testing would be a new feature in 
the part 26 framework. The NRC proposes to permit the use of hair 
testing for only Schedule I or II drugs or their metabolites to inform 
a licensee's or other entity's determination whether the individual is 
trustworthy and reliable. For example, if an individual stated no prior 
use of illegal drugs or potentially addictive habits, a hair screening 
test could be performed during the pre-access process to ascertain the 
validity of the individual's statement. However, if the HHS-certified 
laboratory communicates a laboratory-confirmed positive test result, an 
FFD policy violation may not be administered. This laboratory 
information must be treated as potentially disqualifying FFD 
information, unless the individual subverts the screening process, in 
which case a permanent denial of authorization must be issued under 
proposed Sec.  26.610. To provide assurance of testing effectiveness 
and protections afforded to individuals subject to the FFD program, 
proposed Sec.  26.607(i) would require that an HHS-certified laboratory 
must be used to analyze the hair specimen, a forensic toxicologist must 
review the licensee's or other entity's hair screening process, the 
test kit must be cleared by the FDA, and hair screening must be 
conducted in accordance with the HHS Guidelines. The forensic 
toxicologist review would be necessary if the panel of drug or drug 
metabolites to be tested and their cutoffs are not established by HHS 
or the NRC for hair.
    Proposed Sec.  26.607(j) would allow the use of portal area 
screening for drugs, alcohol, or both. This provision would result in a 
substantial contribution to a licensee or other entity satisfying the 
Sec.  26.23 performance objectives by helping ensure that 100 percent 
of all individuals who arrive at the NRC-licensed facility to perform 
or direct those duties and responsibilities or maintain those types of 
access making them subject to the FFD program are fit for duty and 
deterred from arriving onsite in a physiological condition that may be 
adverse to safety and security. Additionally, screening could be 
conducted when an individual exits the NRC-licensed facility to provide 
assurance that substance abuse had not occurred on the site (see Sec.  
26.23(d)). The screening device could be electronically linked to 
temporarily prevent ingress or egress and could automatically inform 
licensee- or other entity-designated officials of the portal area 
alarm. The proposed requirement would enable the licensee or other 
entity to use innovative technologies to maintain FFD program 
effectiveness when their PMRP compels the licensee or other entity to 
implement mitigative strategies to maintain program effectiveness. The 
use of portal screening technologies may also represent cost savings 
because, for NRC-licensed facilities that have small staff sizes or are 
geographically remote, passive drug and alcohol screening technologies 
could be an innovative alternative to a random testing program, 
although the license or other entity would need to request and receive 
an exemption.
    Proposed Sec.  26.607(j) would also provide that if the portal area 
screening instrument detects a substance that exceeds the instrument's 
established setpoint, the individual would be tested with either a 
collection kit that must be analyzed by an HHS-certified laboratory or 
a POCTA. This situational screening would be equivalent to a for-cause 
test. The requirements would not allow an individual to be rescreened 
by the portal area screening instrument following an initial screening 
detection that exceeded an established setpoint in order to prevent a 
subversion attempt. Similar to other drug and alcohol testing 
technologies enabled for use by subpart M of part 26, a forensic 
toxicology review would be required before using passive screening 
technology to help ensure the effectiveness of the instrument by 
protecting against false positive or negative screening results, which 
would place an unwarranted burden on the individual, licensee, or other 
entity. These instruments and alcohol screening devices, already in the 
marketplace, may also be used to determine true identity to facilitate 
implementation of the FFD BOP, which may be very practicable at 
facilities that operate with small staff sizes.
    Proposed Sec.  26.607(k) would enable the use of a blood specimen 
for drug, alcohol, or other testing for certain medical conditions as 
determined by the licensee- or other entity-designated MRO. This 
requirement would be equivalent to current Sec.  26.31(d)(5). The use 
of a licensee- or other entity-designated MRO and not one designated by 
a third party, such as an MRO

[[Page 86971]]

employed by an offsite specimen collection facility, is important 
because the MRO must be familiar with the subpart M of part 26 
requirements. To help ensure testing effectiveness and protect the 
worker, the blood test would need to be conducted by a laboratory that 
demonstrates compliance with quality control requirements that are 
comparable to those required for certification by the HHS, such as a 
hospital or clinic certified by the State, Commonwealth, or territory.
    Proposed Sec.  26.607(l) would require licensee and other entities 
to use a Federal custody-and-control form (CCF) approved by the OMB for 
the collection and packaging of a hair, oral fluid, or urine specimen. 
This proposed requirement is based on the CCF documentation 
requirements in current subpart E of part 26 because subpart K of part 
26 does not require the use of a CCF under Sec.  26.117(e). 
Additionally, when using a POCTA device, the licensee or other entity 
would be required to implement a licensee- or other entity-approved and 
-maintained procedure that ensures the reliability of the tracking, 
handling, and storage of a specimen from the point of specimen 
collection to final disposition of the specimen and the reliability of 
an identification system to uniquely assign the specimen to the donor. 
Both requirements would help protect the worker by helping ensure chain 
of custody and by contributing to program effectiveness.
    Proposed Sec.  26.607(m) would establish requirements for the 
licensee- or other entity-designated MRO. Section 26.607(m)(1) would be 
equivalent to Sec.  26.405(g), however, the word ``designated'' would 
be added to the first sentence to clarify that the MRO would be 
designated by the licensee or other entity, and not by a third party. 
As stated with regard to proposed Sec.  26.607(k), this change would 
clarify that it is the licensee's or other entity's responsibility, 
through their designated MRO, to determine whether an individual is fit 
for duty and trustworthy and reliable. This would be consistent with 
the description of FFD program personnel in current Sec.  26.31(b) and 
help provide FFD program effectiveness and protections to individuals 
subject to the FFD program. The paragraph was also modified from Sec.  
26.405(g) to address the determinations of FFD policy violations and 
fitness required by subpart H for a part 53 licensee or other entity 
that implements the FFD program described in Sec.  26.605(b).
    Proposed Sec.  26.607(m)(2) would help ensure that MRO reviews are 
consistent with those MRO reviews conducted at other NRC-licensed 
facilities subject to part 26 and that the MRO maintains knowledge of 
drug collection, testing processes and procedures, and evaluation of 
testing results.
    The NRC also proposes that if an MRO performed the duties and 
responsibilities in Sec. Sec.  26.185 and 26.187 for at least three 
continuous years in the last 10 years prior to being hired or 
contracted by the licensee or other entity, then the MRO would not need 
to repeat the initial training and examination requirements. The basis 
for 3 years is that the MRO would have experienced three annual cycles 
of evaluating drug and alcohol test results, contributed to the FFD 
annual report to the NRC, experienced a refueling or maintenance 
outage, understood the duties and responsibilities of individuals 
subject to the FFD program to make informed determinations of fitness, 
demonstrated a safety culture that helps ensure FFD program 
effectiveness, and been subject to NRC inspection. The basis for 10 
years is the relatively long periods between significant changes to 
part 26 and the HHS Guidelines.
    Proposed Sec.  26.607(m)(3) would require that the MRO attend a 
medical- or clinical-based training session on a triennial basis. This 
proposal was developed from Section 13.1 of the HHS Guidelines for 
urine and oral fluid with two substantial differences: the HHS 
Guidelines state that ``requalification training,'' including an exam, 
must be conducted ``at least every 5 years from initial 
certification,'' whereas the proposed Sec.  26.607(m)(3) would require 
a training session every three years. The proposed requirements are 
justified because changes in societal drug use or forensic toxicology 
could occur more frequently than every 5 years, which could compel MROs 
to attend training in areas of forensic toxicology, determinations of 
fitness, or other part 26 technical areas on a more frequent 
periodicity than every 5 years to improve their knowledge and 
expertise.
    Proposed Sec.  26.607(m)(4) would require the MRO to evaluate drug 
testing results by implementing the requirements in Sec.  26.185 or the 
HHS Guidelines through the licensee's or other entity's procedures. 
This requirement would help ensure FFD program effectiveness and 
enhance consistency across the commercial nuclear industry for the 
evaluation of drug testing results. This also would help protect 
individuals because they would be subject to the same evaluation 
criteria. If Sec.  26.185 provides insufficient information for an MRO 
to make a determination on a drug testing result (including adulterant 
and discrepant biological markers), the guidance issued by a State 
agency in the state in which the NRC-licensed facility is located, 
Federal agency, or nationally recognized MRO training and certification 
organization may be used to inform an MRO determination. This provision 
would ensure that the MRO has the flexibility to inform their 
evaluation of the drug testing results and fitness determination, if 
necessary, considering the drug- and alcohol-related flexibilities 
afforded in subpart M of part 26.
    The proposed requirement would also state that an MRO need not 
review a confirmed alcohol positive test result determined by an EBT 
device under Sec.  26.607(c)(3)(vi) and (vii), which are equivalent to 
the current requirements in Sec. Sec.  26.101 and 26.103, respectively. 
The results of an EBT device are precise and accurate enough to support 
the issuance of an FFD policy violation without an MRO review of an EBT 
test result if the instrument demonstrates compliance with the 
requirements in Sec.  26.91. The NRC acknowledges that there are 
physiological conditions that may cause an abnormally high blood 
alcohol concentration, such as diabetes, acid reflux, gastroesophageal 
reflux disease, and perhaps certain diets (high protein and low 
carbohydrates). However, operating experience has not demonstrated a 
compelling need to require an MRO review of all EBT test results. For 
consistency, a licensee or other entity may elect to require its MRO to 
review all EBT test results when a donor communicates a testing concern 
or physiological condition. If the donor has a testing concern, the 
occurrence could be appealed under the proposed Sec.  26.613. If the 
donor presents a physical condition to the MRO that may have caused an 
elevated EBT test result, the MRO may direct an alternative testing 
process (see Sec.  26.607(m)(5)) should it be medically necessary.
    Proposed Sec.  26.607(m)(5) would require the licensee- or other 
entity-designated MRO to determine and approve the use of oral fluid or 
urine as an alternative biological specimen when the donor cannot 
provide a requested specimen for testing. This proposed requirement is 
equivalent to Sec.  26.31(d)(5), which enables the use of an 
alternative specimen collection if a medical condition makes the 
collection of the biological specimen difficult. This determination and 
the retest must be completed as soon as reasonably practicable and 
documented to support recordkeeping, auditing, and NRC inspection.

[[Page 86972]]

    Proposed Sec.  26.607(m)(6) would require that the MRO review all 
specimens screened or tested associated with a drug-related FFD policy 
violation. This includes POCTA, split specimens, and all specimens 
taken to resolve a discrepant condition, such as a possible subversion 
attempt, impairment without a known cause, or a donor-requested or MRO-
directed retest. To resolve a discrepant condition, the MRO is 
authorized to test a specimen for a biological marker, adulterants, or 
additional drugs. The broad scope of this MRO evaluation would be 
necessary because of the variety of different screening and testing 
methods that may have been associated with the FFD policy violation. 
All information learned from the conduct of part 26 drug and alcohol 
screening and testing should be used in the evaluation of an 
individual's trustworthiness and reliability, issuance of a sanction, 
and development of a follow-up treatment and testing plan, if 
administered.
    Proposed Sec.  26.607(n) is equivalent to current Sec.  26.31(d)(6) 
and would establish limits on the screening and testing of biological 
specimens. This is a protection consideration afforded to individuals 
subject to the FFD program and was not provided in subpart K of part 
26. This requirement states that specimens collected under NRC 
regulations may only be designated or approved for screening and 
testing as described in this part and may not be used to conduct any 
other analysis or test without the written permission of the donor. 
Analyses and tests that may not be conducted include, but are not 
limited to, deoxyribonucleic acid (i.e., DNA) testing, serological 
typing, or any other medical or genetic test used for diagnostic or 
specimen identification purposes.
    The NRC proposes to require that no biological specimens may be 
passively sampled and analyzed in a manner different than described in 
subpart M of part 26 to ensure workers are protected from non-
consensual passive screening. The subpart M framework enables passive 
detection of drugs and alcohol, whereas passive detection is not 
afforded in subparts A through I, N, and O of part 26.
    Proposed Sec.  26.607(o) is equivalent to current Sec. Sec.  
26.31(b)(1)(iii)(A) and 26.89 and would require that all specimen 
collections be conducted by a licensee- or other entity-designated and 
-trained individual. For subpart M of part 26, this would include 
onsite specimen collections, except a collection by a portal area 
screening instrument in Sec.  26.607(j).
    Proposed Sec.  26.608 would require licensees and other entities to 
provide FFD program training to individuals subject to the FFD program. 
The proposed performance-based Sec.  26.608 requirement was developed 
from the prescriptive training requirements in current Sec.  26.29 and 
modeled on current Sec.  50.120 and the proposed requirements in 
Sec. Sec.  53.725 and 53.830 because there is no training requirement 
in subpart K of part 26.
    Proposed Sec.  26.608(a)(1) would require an FFD training program 
that includes the licensee's or other entity's FFD policies and 
procedures, including fatigue management, and the individuals' FFD 
program responsibilities. Individuals who collect specimens for testing 
or screening must also be trained in specimen collector duties and 
responsibilities, including, at a minimum, specimen collection, custody 
and control, identification and response to subversion attempts, and 
privacy. The fatigue management training must include the knowledge and 
abilities described in Sec.  26.202(c). For individuals specified in 
Sec.  26.4, a licensee or other entity of a commercial nuclear plant 
would be required to use a SAT as defined in proposed in Sec.  53.725. 
These requirements are based on requirements in Sec.  26.29(a)(2), (3), 
(9), and (10).
    Proposed Sec.  26.608(a)(2) would require training on the BOP. This 
requirement would be based on Sec. Sec.  26.29(a)(8), (9), and (10) and 
26.33. The proposal would require individuals to be trained in the 
detection of behaviors or conditions related to not only illegal drugs, 
as in the current Sec.  26.33 BOP requirements, but also illicit drugs 
and substance abuse onsite and offsite. Also, in reference to 
impairment from fatigue or any cause if left unattended, the phrase in 
Sec.  26.33, ``may constitute a risk to public health and safety or the 
common defense and security,'' would be replaced in Sec.  
26.608(a)(2)(iii) with ``could result in inattentiveness or human 
errors,'' because subpart M of part 26 is focused, in part, on ensuring 
individuals are fit for duty to safely and competently perform or 
direct the performance of assigned duties and responsibilities.
    Proposed Sec.  26.608(a)(2)(iv) focuses on training to inform 
individuals that they are responsible for their own conduct, as well as 
observing others. Specifically, individuals would be trained to 
recognize when they feel unable to safely and competently perform 
assigned duties and responsibilities or act in a trustworthy and 
reliable manner. The proposed training requirement and the proposed 
reporting requirement in Sec.  26.606(a)(5) are in the interest of 
safety and security because the individual is proactively announcing 
that assistance may be necessary. This would be consistent with the 
performance objectives in Sec.  26.23(b) and (c) where certain behavior 
or stress conditions may be indicative of an individual not being fit 
for duty, trustworthy, and reliable.
    Proposed Sec.  26.608(a)(3) would help ensure that individuals 
subject to the FFD program understand that FFD policy violations would 
result in an FFD program sanction and that program information learned 
or generated by FFD program implementation would be used to aide 
licensee or other entity authorization determinations and be shared, as 
requested, with other licensees or other entities subject to parts 26, 
53, and 73. This proposed requirement is equivalent to Sec.  
26.29(a)(1). Proposed Sec.  26.608(a)(3) would be a protection measure 
afforded to individuals subject to the FFD program because they would 
understand that licensees and other entities subject to parts 26, 53, 
and 73 would be informed of, in part, an individual's character, 
reputation, and ability to follow policies, procedures, and 
instructions to safely and competently perform assigned duties and 
responsibilities in a trustworthy and reliable manner. Fitness-for-
duty-related information would include drug and alcohol testing results 
(not quantitative testing values), issuance of any sanctions, FFD-
determinations regarding trustworthiness and reliability, testing 
programs, treatment, and other remedial or corrective action.
    Proposed Sec.  26.608(b) would require individuals be trained and 
receive a trainee assessment before pre-access testing and that 
refresher training and trainee assessments be conducted periodically 
thereafter. These requirements would be equivalent to Sec.  
26.29(c)(1). However, Sec.  26.608(b) was developed from the SAT-based 
training requirements in Sec.  50.120 and training elements from the 
annual training requirements in Sec.  26.29(c)(2). The term ``systems 
approach to training'' would have the meaning in proposed Sec.  
53.725(c). A trainee assessment would be the same as in currently 
required SAT-based training programs.
    Proposed Sec.  26.608(c) would require licensees and other entities 
to periodically evaluate their FFD training programs and revise them as 
appropriate. This training focus is not required by subpart K of part 
26 or Sec.  26.29 but is proposed to address the flexibilities afforded 
in subpart M of

[[Page 86973]]

part 26. This section would be equivalent to Sec.  50.120(b)(3).
    Proposed Sec.  26.609 would require the implementation of a BOP. 
The proposed requirement would be equivalent to that in Sec. Sec.  
26.33 and 26.407, ``Behavioral observation,'' and would apply during 
construction, operation, and decommissioning, if applicable. Because 
subpart M of part 26 would apply during decommissioning through a 
licensee's IMP, proposed Sec.  26.609(a) and (b) were developed, in 
part, from proposed Sec.  73.100(b)(9) and current Sec. Sec.  
73.55(b)(9) and 73.56(f) to help ensure consistency in the conduct of 
behavioral observation whether conducted for FFD or security purposes.
    Under the FFD program, the purpose of the BOP would be to help 
ensure that individuals subject to the FFD program are fit for duty and 
trustworthy and reliable to perform or direct those duties and 
responsibilities and maintain those types of access that make the 
individual subject to the FFD program. This assurance is accomplished 
by requiring each individual subject to subpart M of part 26 to be 
subject to behavioral observation, and by requiring all individuals to 
perform behavioral observation of others and report FFD concerns to the 
licensee- or other entity-designated representative(s). The intent of 
the BOP requirement is not to require that all individuals be observed 
at all times by others; NRC-licensed operators, maintenance 
professionals, security officers, and others routinely perform solo 
operations periodically throughout the day. However, individuals must 
be subject to observation while they are performing or directing the 
performance of duties and responsibilities or maintaining the types of 
access making them subject to the FFD program. Observing behavior only 
at the beginning of a work shift is not sufficient to ascertain whether 
an individual is fit for duty, trustworthy, and reliable. Controlled 
substances may have a delayed effect between use (e.g., ingestion) and 
the onset of physiological or psychological effects, and fatigue 
accumulates with time. Behavior must be continually observed throughout 
the work shift to detect any changes from baseline human performance 
characteristics, including mental or physical health and mannerisms, or 
any activities that may indicate that the individual is not trustworthy 
and reliable.
    Proposed Sec.  26.609(a) would differ from Sec. Sec.  26.33 and 
26.407 in that it would place the responsibility for performing 
behavioral observation on ``all individuals subject to this subpart,'' 
rather than only those ``individuals specified in Sec.  26.4(f) [who] 
are constructing or directing the construction of safety- or security-
related SSCs'' in Sec.  26.407 or ``individuals who are trained under 
Sec.  26.29 to detect behaviors'' in Sec.  26.33 to improve clarity.
    Proposed Sec.  26.609(b) would require all individuals subject to 
the FFD program to report to the licensee- or other entity-designated 
representative any onsite or offsite behaviors or activities by 
individuals subject to this part that may constitute an unreasonable 
risk to the safety or security of the NRC-licensed facility or SNM or 
may cause harm to others. The NRC proposes this description of 
reportable conduct because an individual's activities (e.g., use of 
illegal substances) and communications (e.g., hate speech or threats of 
violence) offsite are a direct indication of the individual's fitness, 
trustworthiness, and reliability and must be evaluated as to whether 
authorization should be granted or maintained. Proposed Sec.  26.609(b) 
would include a description of this conduct instead of the Sec.  26.33 
undefined phrase, ``FFD concerns,'' to enhance the clarity of the 
requirement. This proposed BOP reporting requirement would include any 
information relating to character or reputation of the individual 
indicating that the individual cannot be trusted or relied upon to 
perform those duties and responsibilities or maintain access to NRC-
licensed facilities, SNM, or sensitive information. This would better 
align with the proposed Sec.  73.120 BOP requirement, which states that 
each person subject to behavioral observation must communicate to the 
licensee or applicant observed behaviors or activities of individuals 
that may constitute an unreasonable risk to the health and safety of 
the public and common defense and security. Proposed Sec.  26.609(a) 
and (b) were written broadly to include offsite conduct that the 
reporting individual considers serious enough to call into question the 
character or reputation of the subject individual.
    Proposed Sec.  26.609(c) would require that licensees and other 
entities perform behavioral observation visually, in-person, and, when 
necessary, remotely by live video and audible streaming and capture. 
This requirement was developed from the security observation 
requirements in Sec.  73.55(e)(7)(i)(B) and (C), (h)(2)(v), and (i)(2) 
and (i)(5)(ii). Conducting an in-person observation of another 
individual is the preferred method to ascertain whether the observed 
individual can safely and competently perform assigned duties and 
responsibilities. When in-person observations are not feasible (e.g., 
during solo operations), the proposed requirement would enable the use 
of video monitoring. This is addressed, for example, in proposed Sec.  
26.609(d) regarding NRC-licensed operator manipulation of reactor 
controls. Additionally, certain duties (such as maintenance activities 
performed by a single worker outside of a control room) may not present 
an opportunity for video monitoring; in these situations, behavioral 
observation should be conducted on a sampling basis (i.e., a planned 
observation of the work activity) as outlined in a licensee's or other 
entity's FFD program.
    In situations involving small staff sizes, facilities sited in 
geographically remote locations, or both, additional observers would 
enhance the effectiveness of a BOP. Technological developments in 
automated safety and security systems may enable licensees or other 
entities to reduce staff sizes to 10 to 40 percent of the staff size of 
an LWR facility licensed under part 50 or 52. Smaller staff sizes may 
translate into more solo operations, less teamwork, fewer peer checks, 
or infrequent management oversight of field activities, leading to 
fewer behavioral observations. Therefore, a licensee or other entity 
would have fewer opportunities to observe whether individuals are fit 
for duty. Enabling video and audible streaming and capture to enhance 
the BOP would be consistent with the security-related behavioral 
observation requirement in proposed Sec.  73.120(c)(2)(ii), which would 
also enable video conferencing or other acceptable electronic means 
promoting face-to-face interaction for those individuals working 
remotely.
    Proposed Sec.  26.609(d) would require that licensees or other 
entities perform behavioral observation of NRC-licensed operators who 
manipulate the controls of any commercial nuclear plant licensed under 
part 53, remotely by live video and audible streaming capture for those 
part 53 facilities where individual task loading does not allow for the 
effective conduct of behavior observation in addition to assigned 
operational tasks. The purpose of this paragraph would be similar to 
that of proposed Sec.  26.609(c), where the possibility of in-person 
observation is significantly diminished because of solo operations or 
because the facility may only require a minimum staff size onsite.
    Proposed Sec.  26.610 would be equivalent to Sec.  26.409, 
``Sanctions,'' and would require the licensee or other entity to 
establish sanctions for FFD

[[Page 86974]]

policy violations that, at a minimum, prohibit the individuals 
specified in Sec.  26.4 from being assigned to perform or direct those 
duties and responsibilities or maintaining authorization making them 
subject to subpart M of part 26. To be consistent with Sec.  26.75, 
``Sanctions,'' the severity of the sanction as described in Sec.  
26.610 would escalate with the number of occurrences and severity of 
the FFD policy violation. The sanction would be long enough to help 
deter future FFD policy violations and facilitate counseling and 
treatment before the licensee reinstates the individual's access to the 
facility. The NRC proposes this requirement because the 14-day denial 
described in Sec.  26.75 may not allow sufficient time for counseling 
and treatment based on the particular FFD policy violation.
    Equivalent to Sec.  26.75(c), proposed Sec.  26.610 would also 
require a minimum 5-year denial of access to the NRC-licensed facility 
for certain violations of the FFD policy within the protected area of a 
commercial nuclear plant and by an individual or individuals who are 
the operators of the conveyance to transport or use formula quantities 
of strategic SNM. Equivalent to Sec.  26.75(b), proposed Sec.  26.610 
would require a permanent denial of authorization be issued for any 
subversion attempt.
    Proposed Sec.  26.611 would protect information collected from FFD 
program implementation and would be equivalent to current Sec.  26.411, 
``Protection of information.'' The protected information would include, 
but not be limited to, privacy and medical information. Section 26.611 
would not include the Sec.  26.411 requirement that FFD programs must 
maintain and use the personal information with the highest regard for 
individual privacy because such a requirement would be unnecessary in 
light of the proposed Sec.  26.611(a) requirement that licensees and 
other entities must establish and maintain a system of files and 
procedures to prevent unauthorized disclosure.
    Proposed Sec.  26.611(b), although equivalent to Sec.  26.411(b), 
would require licensees and other entities to have all individuals sign 
a consent to be subject to the FFD program before subjecting the 
individual to the FFD program (e.g., before being subject to a pre-
access test in Sec.  26.607(b)(1), unlike Sec.  26.411(b)). The purpose 
of this proposal would be to enhance protections afforded to 
individuals subject to the FFD program and their knowledge of, in part, 
why they are subject to drug and alcohol testing, behavioral 
observation, information collection, MRO reviews, and other FFD program 
elements. Like the consent required by Sec.  26.411(b), the consent 
would authorize disclosure of the collected information. Consent would 
not be needed for disclosures to the individuals and entities specified 
in Sec.  26.37(b)(1) through (b)(6), (b)(8), and persons deciding 
matters under review in proposed Sec.  26.613, ``Appeals process.''
    Proposed Sec.  26.613 would be equivalent to Sec.  26.413, ``Review 
process.'' The proposed title was changed to an appeal process to 
clarify that Sec.  26.613 would be the process implemented when an 
individual elects to appeal a licensee or other entity determination 
that the individual had violated the FFD policy. The proposal would 
also require that the process include a schedule for the completion of 
the review of the determination that the individual had violated the 
FFD policy. The NRC proposes this requirement because operating 
experience demonstrates that workers may not be protected from a 
continuous review process that does not result in an outcome.
    Proposed Sec.  26.615 would require licensees and other entities to 
perform audits of the FFD program. The proposed section would be 
equivalent to Sec.  26.415, ``Audits.'' Under proposed Sec.  26.615(a), 
audits would be performed at a frequency that ensures the FFD program's 
continuing effectiveness. This would be particularly important for FFD 
program elements that are not part of the FFD PMRP required by Sec.  
26.603(d). Corrective actions would be taken as soon as reasonably 
practicable to resolve any problems identified and preclude recurrence. 
Proposed Sec.  26.615(b) would require the subject matter, scope, and 
frequency of audits be revised as necessary to improve or maintain 
program performance based on findings resulting from licensee or other 
entity implementation of its FFD PMRP. These requirements were 
developed from appendix B to part 50, ``Quality Assurance Criteria for 
Nuclear Power Plants and Fuel Reprocessing Plants''; criterion X, 
``Inspection''; and criterion XVIII, ``Audits.''
    Proposed Sec.  26.615(c) would be equivalent to Sec.  26.415(b) and 
would enable licensees and other entities to conduct joint audits or 
accept audits of C/Vs so long as the audit addresses the relevant 
services of the C/Vs.
    Proposed Sec.  26.615(d) would be equivalent to Sec.  26.415(c) by 
establishing requirements for the auditing of HHS-certified 
laboratories. Unlike Sec.  26.415(c), the proposal would not contain a 
reference to the Department of Transportation drug and alcohol testing 
requirements. This would broaden the regulatory flexibility afforded to 
a licensee or other entity in that they may use an offsite collection 
or testing facility that does not meet the Department of Transportation 
requirements.
    Proposed Sec.  26.615(d) would state that licensees and other 
entities need not audit an HHS-certified laboratory if the licensee's 
or other entity's panel of drugs and drug metabolites to be tested is 
equivalent to the panel by which the laboratory is certified by HHS or 
is subject to the standards and procedures for drug testing and 
evaluation used by the laboratory under the HHS Guidelines. The NRC 
would afford this flexibility because the NRC is aware that HHS desires 
to streamline changes in its guidelines to its panel of drugs and drug 
metabolites to be tested. Therefore, if a licensee or other entity 
elects to implement the HHS Guidelines in its procedures and maintains 
the minimum panel of drugs and drug metabolites to be tested as 
required by subpart M of part 26, a licensee or other entity may still 
use (and not audit) the HHS-certified laboratory because the Sec.  
26.603(e) change control process would maintain FFD program 
effectiveness.
    To help ensure FFD program effectiveness, Sec.  26.615(d) would 
also require that collection facility procedures are comparable to 
those required in subpart E of part 26, including a proposed 
requirement that the offsite facility's specimen collection and testing 
procedures are audited on a biennial basis, which is also a protection 
consideration afforded to individuals subject to the FFD program. 
Conducting this audit on a biennial basis would be equivalent to that 
required in Sec.  26.41(b) and would help ensure that the specimen 
collection process at the facility remains effective.
    Proposed Sec.  26.617 would establish recordkeeping and reporting 
requirements equivalent to those in current Sec.  26.417. However, 
Sec.  26.617 would require retention of records pertaining to 
administration of the FFD program and FFD performance data required by 
Sec.  26.717 until license termination, which is based on current Sec.  
26.711(a) because Sec.  26.417 does not provide for a retention period.
    Proposed Sec.  26.617(b)(1) would be identical to the reporting 
requirements in Sec.  26.417(b)(1) regarding the licensee's or other 
entity's FFD program.
    Proposed Sec.  26.617(b)(2) would require the reporting of annual 
(i.e., January through December) program performance information to the 
NRC before March 1 of the following year. This reporting would be 
equivalent to

[[Page 86975]]

the annual program performance requirement in Sec.  26.417(b)(1), and 
the March 1 due date is based on the reporting deadline in Sec.  
26.717(e). Licensees and other entities would be required to report FFD 
performance information using new NRC Forms 893, ``Single FFD Policy 
Violation Form,'' and 894, ``10 CFR part 26, subpart M, Annual 
Reporting Form for FFD Performance Information.''
    Proposed Sec.  26.617(c) would require that FFD-related information 
be shared within the commercial nuclear industry when requested to 
support authorization determinations. This requirement would help 
individuals seeking employment by another NRC-licensed facility subject 
to subpart C of part 26, complete their NRC-required sanctions and 
licensee-administered or -directed drug and/or alcohol abuse treatment 
plans before the restoration of authorization by a licensee or other 
entity. Information sharing may also enhance FFD program effectiveness 
because FFD-related lessons learned from, for example, substance 
testing, subversion attempts, and laboratory and MRO performance must 
be shared when requested.
    Proposed Sec.  26.619 would require licensees or other entities to 
establish a process to evaluate individuals when their fitness or 
trustworthiness and reliability are in question. Section 26.619 would 
be equivalent to Sec.  26.419, ``Suitability and fitness 
determinations,'' but, unlike Sec.  26.419, would apply during the 
construction and operation phases. Also, proposed Sec.  26.619 would 
require that a suitability or fitness determination conducted for cause 
be conducted face-to-face. This proposed requirement is based on 
current Sec.  26.189(c); however, unlike Sec.  26.189(c), proposed 
Sec.  26.619 would not prohibit augmenting determinations via 
electronic means of communication. Instead, Sec.  26.619 would 
explicitly permit determinations to be performed via electronic means, 
so long as those determinations are supported by an appropriately 
trained individual who is present in-person with the individual being 
assessed.
    In considering the current restriction on the use of electronic 
means of communication for determinations of fitness conducted for 
cause, the NRC finds that since publication of the 2008 part 26 final 
rule, there have been developments in using electronic means of 
communication (i.e., ``videoconferencing'') as an alternative to 
conducting face-to-face interactions. To address these considerations, 
the NRC contracted the Pacific Northwest National Laboratory (PNNL), 
DOE, to study whether a medical and mental health assessment via 
electronic communication could be an acceptable alternative to an in-
person, face-to-face assessment.\13\ Based on this study, if electronic 
means were to be used to conduct a face-to-face assessment, an in-
person element would still be integral to the assessment process. 
However, under certain circumstances, face-to-face determinations and 
assessments conducted as part of an FFD program for an entity licensed 
under part 53 (i.e., those determinations and assessments performed in 
accordance with Sec.  26.619, Sec.  26.207, or Sec.  26.211) may be 
augmented via electronic communications. Such remotely conducted 
determinations and assessments would be required to be conducted with 
someone who is present in-person with the individual being assessed and 
who is trained in accordance with the requirements of either Sec.  
26.29 and Sec.  26.203(c) or Sec.  26.608 and Sec.  26.202(c). 
Permitting the use of electronic communications would help ensure FFD 
program effectiveness, especially in instances where the part 53 
commercial nuclear plant is sited in a geographically remote location 
or when the facility has a small staff size.
---------------------------------------------------------------------------

    \13\ PNNL, Technical Letter Report, ``The Use of Electronic 
Communications to Perform Determinations of Fitness,'' dated August 
2017.
---------------------------------------------------------------------------

D. Proposed Changes to Part 26, Subpart N

    Proposed Sec.  26.709 would make the recordkeeping and reporting 
requirements in subpart N of part 26 applicable to licensees and other 
entities of facilities licensed under part 53 that elect not to 
implement the requirements in subpart M of part 26 or elect to 
implement the requirements in Sec.  26.605(b).
    Proposed Sec.  26.711(c) and (d) would be amended to make these 
requirements applicable to licensees or other entities described in 
Sec.  26.3(f). Section 26.711(c) provides protection to individuals 
subject to part 26 by enabling an individual's right to review FFD-
related information and correct any inaccurate or incomplete 
information. Section 26.711(d) requires, in part, that any FFD-related 
information shared with other licensees or other entities is correct 
and complete.

E. Proposed Changes to Part 26, Subpart O

    The vast majority of the proposed changes to part 26 would be new 
or revised substantive provisions that would establish a regulatory 
obligation or prohibition or would be conforming edits to reflect the 
addition of part 53. The only new provision that would not be 
substantive, such that violation of it would not result in a criminal 
penalty, would be proposed Sec.  26.601. Therefore, the NRC proposes to 
add Sec.  26.601 to the list of regulations in Sec.  26.825(b) to which 
criminal sanctions do not apply.

10 CFR Part 50

A. Section 50.160: Emergency Preparedness for Small Modular Reactors, 
Non-Light-Water Reactors, and Non-Power Production or Utilization 
Facilities

    This proposed rule would revise Sec.  50.160(b)(3) and (c)(2) to 
make that section applicable to applicants and licensees under part 53. 
Section 50.160 provides an alternative to other part 50 emergency 
preparedness requirements focused on large light-water reactors to 
provide an optional emergency preparedness framework specifically for 
small modular reactors (SMRs) and other new technologies. These 
alternative emergency preparedness requirements adopt a performance-
based, technology-inclusive, risk-informed, and consequence-oriented 
approach. Commercial nuclear reactor applicants complying with Sec.  
50.160 would be required to submit as part of the application the 
analysis used to determine whether the criteria in Sec.  
53.1109(g)(2)(i)(A) and (B) are met and, if they are met, the size of 
the plume exposure pathway emergency planning zone (EPZ). An EPZ bounds 
the area surrounding a facility within which detailed planning is 
needed to implement predetermined, prompt protective actions. The 
criterion in proposed Sec.  53.1109(g)(2)(i)(A) is that public dose, as 
defined in Sec.  20.1003, is projected to exceed 10 mSv (1 rem) TEDE 
over 96 hours from the release of radioactive materials from the 
facility considering accident likelihood and source term, timing of the 
accident sequence, and meteorology. The criterion in proposed Sec.  
53.1109(g)(2)(i)(B) is that pre-determined, prompt protective measures 
are necessary. These are the same criteria that are in Sec.  
50.33(g)(2)(i)(A) and (B) and are used to assess the need for and size 
of an EPZ in applications under parts 50 and 52.
    Applicants choosing to comply with Sec.  50.160 must determine the 
radiological releases from the facility that are evaluated in the 
determination of the plume exposure pathway EPZ. Consistent with other 
Federal guidelines such as the Federal Emergency Management Agency 
``Radiological Emergency Preparedness Program Manual,'' issued in 2023, 
and the

[[Page 86976]]

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

B. Appendix B to Part 50: Quality Assurance Criteria for Nuclear Power 
Plants and Fuel Reprocessing Plants

    Appendix B to part 50 would be amended to make it applicable to 
applicants and licensees under part 53. This results in the need for 
some revisions to recognize differences in terminology between parts 50 
and 53. Namely, the term ``design bases,'' which is defined in Sec.  
50.2, is not used in part 53. For this reason, text is added in both 
Section III, ``Design Control,'' and Section IV, ``Procurement Document 
Control,'' to refer to ``functional design criteria, as defined in 
Sec.  53.020,'' as the part 53 equivalent of the term ``design bases.''

10 CFR Part 73

A. Section 73.100: Technology-Inclusive Requirements for Physical 
Protection of Licensed Activities at Commercial Nuclear Plants Against 
Radiological Sabotage

    Proposed Sec.  73.100 would provide a performance-based regulatory 
framework for the design, implementation, and maintenance of a physical 
protection program and security organization for certain commercial 
nuclear plants licensed under part 53. The current Sec.  73.55 physical 
security requirements for nuclear power reactors licensed under part 50 
and part 52 use a combination of performance criteria (e.g., Sec.  
73.55(b)(1) through (3)) and numerous prescriptive requirements 
developed to achieve performance objectives (e.g., Sec.  
73.55(k)(5)(ii)). By contrast, in the proposed performance-based 
approach to physical security for part 53, performance objectives and 
requirements would be the primary bases for regulatory decision-making, 
giving the licensee the flexibility to determine how to demonstrate 
compliance with the established performance criteria for an effective 
physical protection program. This proposed physical protection program 
would provide an optional pathway for licensees that elect not to 
demonstrate compliance with the provisions in Sec.  73.55 and do not 
satisfy the criterion as described in proposed Sec.  53.860(a)(2). This 
proposed physical protection program would provide that activities 
involving SNM are not inimical to the common defense and security and 
do not constitute an unreasonable risk to the public health and safety.
    Section 73.100(a) would require each part 53 licensee that elects 
to demonstrate compliance with this section rather than Sec.  73.55 to 
implement the requirements therein through its physical security plan, 
training and qualification plan, safeguards contingency plan, and 
cybersecurity plan (referred to collectively hereafter as ``security 
plans'') prior to initial fuel load into the reactor (or, for a fueled 
manufactured reactor, before initiating the physical removal of any one 
of the independent physical mechanisms to prevent criticality required 
under Sec.  53.620(d)(1)). The security plans would need to identify, 
describe, and account for site-specific conditions that affect the 
licensee's capability to satisfy the requirements of Sec.  73.100. 
Based on experience from recent new reactor licensing reviews, the NRC 
recognizes that licensees may seek to receive unirradiated fuel onsite 
before carrying out the security requirements in Sec.  73.100. However, 
these security requirements would have to be implemented at some point 
before reactor operation to address the increased risk arising from 
irradiated fuel onsite. This proposed rule would make clear that part 
53 applicants and licensees using Sec.  73.100 may bring unirradiated 
nuclear fuel onsite and protect it in accordance with the NRC's 
requirements for physical protection of SNM of moderate and low 
strategic significance under Sec.  73.67 until initial fuel load into 
the reactor (or, for a fueled manufactured reactor, until initiating 
the physical removal of any one of the independent physical mechanisms 
to prevent criticality required under Sec.  53.620(d)(1)).
    Section 73.100(b) would outline the general performance objective 
and design requirements of the licensee physical protection program. A 
licensee's program would be required to provide protection against any 
deliberate act within the DBT of radiological sabotage, including spent 
fuel sabotage, which could directly or indirectly endanger the public 
health and safety by exposure to radiation. The physical protection 
program is supported by the AA program, cybersecurity program, and IMP 
to demonstrate compliance with the general performance objective of 
Sec.  73.100(b).
    Section 73.100(b)(2) was developed, in part, from Sec.  
73.55(b)(3). To satisfy the general performance objective of Sec.  
73.100(b)(1), the physical protection program would need to protect 
against the DBT of radiological sabotage. The existing fleet of LWR 
satisfies this objective by preventing significant core

[[Page 86977]]

damage and spent fuel sabotage. Some non-LWR reactor licensees' 
physical protection programs may be designed to prevent a significant 
release of radionuclides from any source. Therefore, the proposed 
performance objective would focus on radiological sabotage in general, 
rather than a specific focus on core damage or spent fuel sabotage, to 
be technology inclusive and allow for flexibility for different reactor 
technologies.
    Under the proposed Sec.  73.100(b)(2)(ii), licensees must provide 
defense in depth in achieving performance requirements through the 
integration of engineered systems, administrative controls, and 
management measures. This requirement would apply defense-in-depth 
concepts as part of the physical protection program to ensure the 
capability to demonstrate compliance with the performance objective of 
the proposed Sec.  73.100(b)(1) is maintained in the changing threat 
environment. The defense-in-depth philosophy applies to measures 
against intentional acts as required by Sec.  73.100(b), and the 
designs of physical security systems should employ defense in depth 
through systems diversity, independence, and separation under Sec.  
73.100(b)(2). The most common defense-in-depth measures apply concepts 
of redundancy, diversity, independence, and safety margin to ensure 
systems reliability and availability. The defense-in-depth philosophy 
applies to the design of a physical protection program, which 
integrates engineered controls and administrative controls, to provide 
protection against the DBT for radiological sabotage.
    Section 73.100(b)(3) would require the physical protection program 
to be designed and implemented to achieve and maintain the reliability 
and availability of SSCs required for demonstrating compliance with 
specified performance requirements. These physical protection 
performance requirements were informed by Sec.  73.55(b) and the 
Commission's Advanced Reactor Policy Statement.
    The performance objective of protecting against the DBT of 
radiological sabotage is achieved by the design and implementation of 
the physical protection program, maintained at all times, with the 
following required performance capabilities proposed in the provisions 
in Sec.  73.100(b)(3): intrusion detection, intrusion assessment, 
security communication, security response, protecting against land and 
waterborne vehicle bomb assaults, and access control portals. The 
physical protection program must maintain the reliability and 
availability of SSCs relied upon for demonstrating compliance with the 
performance requirements. The terms ``reliability and availability'' 
are intended to describe defense in depth in a performance-based manner 
and would be critical elements for demonstrating compliance with the 
proposed requirement for protection against the DBT of radiological 
sabotage as described in the proposed Sec.  73.100(b)(2).
    The first element, ``intrusion detection,'' would be provided 
through the use of detection equipment, patrols, access controls, and 
other program elements and would provide notification to the licensee 
that a potential threat is present and where the threat is located.
    The second element, ``intrusion assessment,'' would provide a 
mechanism through which the licensee would identify the nature of the 
threat detected. This would be accomplished through the use of video 
equipment, patrols, and other program elements that would provide the 
licensee with timely information about the threat for use in 
determining how to respond.
    The third element, ``security communication,'' would provide a 
mechanism through which the licensee would communicate the necessary 
information to the response force to ensure effectiveness of the 
physical protection program. This would be accomplished through the 
redundant, independent, and diverse design of physical security and/or 
plant SSCs relied on for onsite and offsite security communications. 
The continuity and integrity of communications should account for the 
DBT's ability to affect the reliability and availability of 
communications.
    The fourth element, ``security response,'' would provide a 
mechanism through which the licensee would be capable of timely 
security response to interdict and neutralize threats up to and 
including the DBT of radiological sabotage. The security response may 
include the use of onsite armed responders, law enforcement responders 
(local, State, or Federal), or other offsite armed responders (e.g., 
licensee proprietary or contract security personnel who are positioned 
offsite), or a combination thereof, as appropriate.\14\ The licensee 
must provide protection against any element of the DBT, to include 
those that do not rise to the full capability of the DBT. Structures, 
systems, and components relied on to provide delay functions must be 
designed to provide for timely response to adversary attacks with 
adequate defense in depth. Delay would allow the licensee to take 
necessary actions to counter any attempt by the threat to advance 
towards the protected target or target set element. The overall 
response objective would be to place the threat in a condition from 
which the threat no longer has the potential for, or capability of, 
doing harm to the protected target.
---------------------------------------------------------------------------

    \14\ The NRC's security regulations for commercial nuclear power 
reactors have historically considered onsite armed responders to be 
the only acceptable method for interdicting and neutralizing threats 
up to and including the DBT of radiological sabotage. The proposed 
rule would permit advanced power reactor licensees to use any 
interdiction and neutralization method, which would be an extension 
of the Commission's position in SRM-SECY-17-0100, ``Security 
Baseline Inspection Program Assessment Results and Recommendations 
for Program Efficiencies,'' dated October 8, 2018, and SRM-SECY-20-
0070, ``Technical Evaluation of the Security Bounding Time Concept 
for Operating Nuclear Power Plants,'' dated June 6, 2024. Under the 
proposed rule, a licensee would retain the responsibility to detect, 
assess, interdict, and neutralize threats up to and including the 
DBT of radiological sabotage, but would be able to rely on law 
enforcement or other offsite armed responders as a method for 
fulfilling the required interdiction and neutralization 
capabilities. For licensees that choose to rely on law enforcement 
to fulfill these capabilities, the proposed rule would not create 
any NRC regulatory jurisdiction over, or requirements for, law 
enforcement. In SRM-SECY-23-0021, ``Proposed Rule: Risk-Informed, 
Technology-Inclusive Regulatory Framework for Advanced Reactors (RIN 
3150-AK31),'' dated March 4, 2024, the Commission approved a similar 
approach to defend against radiological sabotage.
---------------------------------------------------------------------------

    The fifth element, ``protecting against land and waterborne vehicle 
bomb assaults,'' would provide a mechanism through which the licensee 
would be capable of protecting the plant against the DBT vehicle bomb 
assault. The methods that are relied on to protect against a DBT land 
vehicle and waterborne vehicle bomb assault must be designed to protect 
the reactor building, structures containing safety or security related 
systems, and components from explosive effects.
    The sixth element, ``access control portals,'' would provide a 
mechanism through which the licensee would be capable of detecting and 
denying unauthorized access to persons and pass-through of contraband 
materials (e.g., weapons, incendiaries, explosives) to protected areas. 
Integrity of the access control system is maintained through licensee 
oversight and ensures that attempts to circumvent or bypass the 
established process will be detected and access denied.
    The proposed performance requirements would permit the applicant or 
licensee to determine how to design the physical protection program to 
protect the plant against the DBT of radiological sabotage without

[[Page 86978]]

prescriptive requirements such as those currently found in Sec.  73.55. 
DG-5076, ``Guidance for Technology Inclusive Requirements for Physical 
Protection of Licensed Activities at Commercial Nuclear Plants,'' has 
been developed by the NRC to describe one acceptable approach to 
demonstrate compliance with requirements proposed in Sec.  73.100.
    Section 73.100(b)(4) would require the licensee to identify target 
sets in accordance with Sec.  73.55(f). For non-LWR and SMRs, target 
sets would be defined in DG-5071, ``Target Set Identification and 
Development for Nuclear Power Plants,'' as the minimum combination of 
equipment, operator actions, and/or structures that, if all are 
prevented from performing their intended safety function or prevented 
from being accomplished, barring extraordinary actions by plant 
operations, would likely result in a significant release of 
radionuclides from any source (e.g., a release to the environment 
exceeding that analyzed in the DBA licensing basis).
    Section 73.100(b)(5) would require that each licensee perform a 
site-specific analysis for the purpose of identifying and analyzing 
site-specific conditions that affect the design of the onsite physical 
protection program.
    Section 73.100(b)(6) would require licensees to implement a 
performance evaluation program, which would ensure that a licensee will 
periodically test and evaluate the effectiveness of the physical 
protection program to protect against the DBT. This program would 
ensure that licensees are able to demonstrate that the physical 
protection program satisfies the response requirements of Sec.  73.100 
and that the site's protective strategy effectively protects against 
the DBT. Licensee performance evaluations would include methods to 
assess, test, and challenge the integration of the physical protection 
programs functions and demonstrate the effectiveness of security plans, 
licensee protective strategy, and implementing procedures in accordance 
with Sec.  73.100(g).
    Section 73.100(b)(7) would require licensees to implement an AA 
program in accordance with Sec.  73.56. Section 73.100(b)(8) would 
require licensees to establish, maintain, and implement protection 
against a cyberattack based on either the proposed cybersecurity 
program described in Sec.  73.110 or the program described in existing 
Sec.  73.54.
    Section 73.100(b)(9) would require an IMP that monitors the initial 
and continuing trustworthiness and reliability of individuals granted 
or retaining unescorted access or unescorted AA to a protected or vital 
area. The IMP must also implement defense-in-depth methodologies to 
minimize the potential for an insider (active, passive, or both) to 
adversely affect the licensee's capability to protect against 
radiological sabotage. Because no one element of the AA program, FFD 
program, cybersecurity program, or physical protection program, would, 
by itself, provide the level of protection against the insider 
necessary to demonstrate compliance with the performance objective of 
the proposed Sec.  73.100(b), the effective integration of these 
programs is a necessary requirement to achieve defense in depth against 
the potential insider.
    Section 73.100(b)(10) would require that the licensee have the 
capability to track, trend, correct, and prevent recurrence of failures 
and deficiencies in the implementation of the requirements of this 
section. Section 73.100(b)(11) would require the coordination of the 
security plans and associated procedures with other onsite plans to 
manage the safety and security interface during normal or emergency 
operations.
    Section 73.100(c) was developed from Sec.  73.55(c)(7), ``Security 
implementing procedures,'' and Sec.  73.55(d), ``Security 
organization,'' and would outline the requirements for the composition, 
equipping, and training of the security organization. The purpose of 
the security organization is to effectively implement the physical 
protection program. Individuals assigned to perform physical protection 
or contingency response duties must be trained, equipped, and qualified 
to perform assigned duties and responsibilities.
    Section 73.100(d) would establish a performance requirement for 
searches of personnel, vehicles, and materials for the protection 
against radiological sabotage. The requirement describes broad 
categories of material (explosives, firearms, incendiary devices, etc.) 
to be detected and prevented from entry into the protected area; 
specific items that will be prohibited would not be prescribed in the 
regulation but will be stated in the licensee security plans with 
detailed descriptions being identified in implementation procedures.
    Section 73.100(e) would require a training and qualification 
program, described in the training and qualification plan, that ensures 
personnel are able to effectively perform their assigned security-
related job duties. This high-level requirement would allow flexibility 
in how the licensee chooses to train its security personnel. One method 
for accomplishing this requirement would be to provide a training and 
qualification program that would be equivalent to appendix B to part 
73.
    Section 73.100(f) would require periodic security reviews of the 
physical protection program to ensure effective implementation of the 
program by independent individuals. The evaluation process would 
provide a systematized approach for assessing the physical protection 
program as a basis for further development and improvement. Program 
reviews should be designed to ensure that the physical protection 
program maintains effectiveness and demonstrates compliance with NRC 
requirements. Section 73.100(f)(1) was developed from Sec.  73.55(m) 
and would require review of each element of the physical protection 
program. Section 73.100(f)(2) would require licensees to perform self-
assessments of physical protection program functions to ensure that the 
capability to detect, assess, interdict, and neutralize the DBT of 
radiological sabotage is maintained. Section 73.100(f)(3) would require 
an audit of the effectiveness of the physical protection program; 
security plans; implementing procedures; cybersecurity programs; 
management of the safety/security interface activities; the testing, 
maintenance, and calibration program; and response commitments by 
local, State, and Federal law enforcement authorities. Section 
73.100(f)(4) would require that results and recommendations, management 
findings, and any actions taken be documented and maintained to be 
available for inspection by the NRC. These reviews are independent of 
the ongoing performance evaluations described in Sec.  73.100(b)(6) and 
(g).
    Section 73.100(g) would require that licensee performance 
evaluations, described in Sec.  73.100(b)(6), include methods 
appropriate and necessary to assess, test, and challenge the 
integration of the physical protection program's functions to protect 
against the DBT. The performance evaluations must also address the 
licensee's measures to protect against cyberattacks, in accordance with 
the required cybersecurity plan, and engineered systems designed to 
protect against the DBT standalone ground vehicle bomb attack.
    Section 73.100(h) would establish performance requirements for 
maintaining security SSCs relied on to perform security functions to 
protect against the DBT. It would require that corrective actions and 
compensatory measures be taken by a licensee in response to a 
degradation of security

[[Page 86979]]

equipment or failure of the equipment to perform its intended 
functions. The licensee would be required to maintain the SSCs 
described in its design and licensing basis to ensure that they are 
reliable and available.
    Section 73.100(i) would establish requirements for the suspension 
of security measures in response to emergency and extraordinary 
conditions. The requirements of this paragraph, which were developed 
from Sec.  73.55(p), would be intended to provide flexibility to a 
licensee for taking reasonable actions that depart from a security plan 
in an emergency when such actions are immediately needed to protect the 
public health and safety and no action consistent with license 
conditions and TS that can provide adequate or equivalent protection is 
immediately apparent in accordance with proposed Sec.  53.740(h).
    Section 73.100(j) would establish requirements regarding the 
inspection, retention and maintenance of records required to be kept by 
the NRC regulations, orders, or license conditions. These proposed 
requirements are developed from Sec.  73.55(q).

B. Section 73.110: Technology-Inclusive Requirements for Protection of 
Digital Computer and Communication Systems and Networks

    Section 53.860 would require that a licensee establish, implement, 
and maintain a cybersecurity program in accordance with Sec.  73.54 or 
Sec.  73.110. Section 73.110 would establish requirements for the 
development and maintenance of a cybersecurity program for commercial 
nuclear plants licensed under part 53. This proposed section would 
implement a graded approach to determine the level of cybersecurity 
protection required for digital computers, communication systems, and 
networks. The proposed new section is informed by: (1) the operating 
experience from power reactors and fuel cycle facilities; and (2) the 
existing Sec.  73.54 framework, which addresses some of the basic 
issues for cybersecurity regardless of the type of reactor. Differences 
between the Sec.  73.54 requirements and those proposed in Sec.  73.110 
are primarily based on the implementation of a consequence-based 
approach to cybersecurity that provides flexibility to accommodate the 
wide range of reactor technologies to be assessed by the NRC. A graded 
approach based on consequences is intended to account for the differing 
risk levels among reactor technologies. Specifically, the proposed new 
section would require licensees to demonstrate protection against 
cyberattacks in a manner that is commensurate with the potential 
consequences from those attacks.
    Under proposed Sec.  73.110(a), licensees would need to ensure that 
digital computer and communications systems are adequately protected 
against a potential cyberattack that would result in: (1) a scenario 
where the cyberattack leads to offsite radiation doses that would 
endanger public health and safety (i.e., the resulting consequence 
exceeds the reference dose values in Sec.  53.210); or (2) a scenario 
where the cyberattack adversely impacts the physical security digital 
assets used by the licensee to prevent unauthorized removal of material 
or radiological sabotage. Security digital assets would include those 
used for nuclear MC&A.
    The proposed Sec.  73.110(b) would require licensees to protect the 
communication system and networks associated with the functions 
described in Sec.  73.110(a)(1) and (a)(2) from cyberattacks. To 
accomplish this, the licensee would establish, implement, and maintain 
a cybersecurity program for protecting digital assets within the scope 
of Sec.  73.110 that would make use of risk insights, including threat 
information, and would consider the resulting level of consequences of 
the threats. If the outcome of the assessment by the licensee under 
Sec.  73.110(b)(1) revealed that a potential cyberattack would not 
compromise any digital assets that support safety and security 
functions, and thus would not result in the consequences listed in 
Sec.  73.110(a) (e.g., would not exceed the reference dose values), 
then only a narrow set of the cybersecurity program requirements in 
Sec.  73.110(d) and (e) would apply. For example, the licensee would 
only need to develop a cybersecurity program that implements the 
requirements dealing with:
     Analyzing modifications of any asset before implementation 
to see if they demonstrate compliance with the potential consequences 
in Sec.  73.110(a);
     Ensuring employees and contractors are aware of 
cybersecurity requirements and have some level of cybersecurity 
training;
     Evaluating and managing cybersecurity risks to the plant;
     Reviewing the cybersecurity plan for any required changes; 
and,
     Retaining records of the cybersecurity plan along with any 
plan changes.
    Section 73.110(c) through (e) were developed from Sec.  
73.54(a)(2), and (c) through (h), respectively.
    The proposed requirements would address the need for the licensee 
to develop a cybersecurity program that implements a defense-in-depth 
protective strategy as required by proposed section Sec.  73.110(d)(2). 
A defense-in-depth protective strategy for cybersecurity is represented 
by collections of complementary and redundant security controls that 
establish multiple layers of protection to safeguard critical digital 
assets. Under a defense-in-depth protective strategy, the failure of a 
single protective strategy or security control should not result in the 
compromise of safety and security functions.

C. Section 73.120: Access Authorization Program for Commercial Nuclear 
Plants

    Section 73.120 would address AA for certain commercial nuclear 
plants licensed under part 53. The proposed language in Sec.  73.120 
would provide an alternate approach to the existing framework for AA 
under Sec. Sec.  73.55, 73.56, and 73.57, commensurate with risk and 
consequences to public health and safety. It would be available to part 
53 applicants and licensees who demonstrate in an analysis that the 
offsite consequences of a DBE satisfy the criterion defined in Sec.  
53.860(a)(2)(i) (i.e., would not exceed the offsite dose values in 
Sec.  53.210(b)). The proposed requirements in Sec.  73.120 would be 
similar to the existing AA program elements for those NRC licensed 
facilities issued additional security measures (ASMs) orders and for 
materials licensees under Sec.  37.21. Applicants not satisfying the 
criterion would need to establish, implement, and maintain a full AA 
program, including an IMP, in accordance with Sec.  73.56.
    Proposed Sec.  73.120(a) would be based on an applicant satisfying 
the eligibility criterion in Sec.  53.860(a)(2)(i). Section 73.120(b) 
would identify the categories of individuals who would be subject to an 
AA program in accordance with this section. The applicability statement 
in Sec.  73.120(b)(1)(i) would encompass individuals whom the licensee 
intends to grant unescorted access to the facilities' most sensitive 
areas, consistent with Sec.  73.56(b)(1)(i) for power reactors and the 
ASM orders and license conditions issued to any NRC licensed facility 
or material licensee. Sections 73.120(b)(1)(ii) through (iv) would be 
consistent with Sec.  73.56(b)(1)(ii) through (iv), respectively. The 
program would include individuals who may be onsite or offsite (e.g., 
remote operators or information technology staff) and have virtual 
access to important plant operational and communication systems based 
upon assigned duties and

[[Page 86980]]

responsibilities. An individual who has remote access to plant 
equipment and communication systems may have trusted privileges greater 
than the personnel at the plant site. Section 73.120(b)(1)(iii) would 
state that offsite law enforcement personnel on official duty would not 
be subject to the licensee AA program.
    Section 73.120(c) would provide general performance objectives and 
requirements largely consistent with the AA program requirements for 
nuclear power reactors under Sec.  73.56 and would provide licensees 
and applicants the flexibility in establishing their AA program to 
demonstrate compliance with various performance objectives.
    Section 73.120(c)(1) would include background investigation 
requirements consistent with Sec.  37.25, as well as ASMs and license 
conditions that are applied to non-power reactor licensees. Background 
investigations include important elements to establish the 
trustworthiness and reliability of an individual, such that they do not 
constitute an unreasonable risk to public health and safety or the 
common defense and security. These include the following: (1) personal 
history disclosure, (2) verification of true identity, (3) employment 
history evaluation, (4) unemployment/military service/education, (5) 
credit history evaluation, (6) character and reputation evaluation, and 
(7) Federal Bureau of Investigation criminal history record check.
    Section Sec.  73.120(c)(2) would establish behavioral observation 
requirements, which are an awareness initiative for recognizing 
behaviors adverse to the safe operation and security of the facility 
through observing the behavior of others in the workplace and reporting 
aberrant behavior or changes in behavior that might reflect negatively 
on an individual's trustworthiness or reliability. Maintaining 
behavioral observation would assist and/or improve worker safety and 
reduce the risk of an insider threat. This proposed requirement in 
Sec.  73.120(c)(2) would be a scaled version of the full BOP required 
under Sec.  73.56(f).
    Section Sec.  73.120(c)(2) would provide licensees greater 
flexibility to implement behavioral observation options for individuals 
granted unescorted access to the commercial nuclear plant's protected 
area. Such options on reporting questionable behavior may include a 
program similar to the Department of Homeland Security's program, ``If 
you see something, say something,'' or to a corporate behavioral 
awareness program. Commensurate with the potential lower safety and 
security risks of a commercial nuclear plant that meets the criterion 
in Sec.  53.860(a)(2)(i), Sec.  73.120(c)(2) would not require the 
establishment of a comprehensive training program for behavioral 
observation (i.e., initial and refresher training including knowledge 
checks) as required for power reactors under Sec.  73.56 and part 26. 
Under Sec.  73.120(c)(2)(ii), behavioral observation would be able to 
be performed in-person or remotely by video, and identified behavior of 
concern would need to be reported to plant supervision. The remote 
access alternative to face-to-face interactions provides substantial 
flexibility for licensees and applicants. Any video conferencing or 
other acceptable electronic means promoting face-to-face interaction 
for those individuals working remotely would demonstrate compliance 
with this regulation.
    Section 73.120(c)(3) captures and maintains the self-reporting of 
legal actions as an essential performance element to enhance the 
licensee's behavioral observation initiative similar to the current 
requirements under Sec.  73.56(g), assuring that personnel who are 
granted and who maintain unescorted access are trustworthy and 
reliable.
    Section 73.120(c)(4) would provide a scalable approach for granting 
and maintaining unescorted access. One component not included from 
Sec.  73.56 is the need for a psychological assessment and reassessment 
under Sec.  73.56(e) for granting unescorted access and Sec.  
73.56(i)(v)(B) for individuals who perform one or more of the job 
functions described in Sec.  73.120(b)(1)(ii) for maintaining 
unescorted access. Moreover, the requirement would permit criminal 
history updates to be completed within 10 years of the last review, 
compared to the three- or five-year reinvestigation periodicity for 
personnel at an operating commercial nuclear plant. In addition, no 
credit check re-evaluation would be required for these individuals.
    The continued need to maintain unescorted access would be evaluated 
on an annual basis by the reviewing official. Guidance in DG-5074, 
``Access Authorization Program for Commercial Nuclear Plants,'' would 
specify that this evaluation should be based on a compilation of 
personnel interactions as described in the licensee's or applicant's 
policy and procedures for behavioral observation and the maintenance of 
an approved AA list.
    Section 73.120(c)(5) would require licensees and applicants to 
determine when a person no longer requires the need for unescorted 
access or no longer satisfies the AA requirement found within this 
section. Guidance in DG-5074 would further explain that licensees have 
the flexibility to terminate unescorted access to specific areas of the 
site if individuals lack the continued need for that access to perform 
their duties and responsibilities.
    Section 73.120(c)(6) would be consistent with the purpose of Sec.  
37.23(e) and would include the individual's right to correct and 
complete information as required under Sec.  37.23(g). The section 
would include a requirement for designating a reviewing official. The 
language would provide clarity regarding the roles and responsibility 
of a reviewing official, who would be the only individual authorized to 
make unescorted access determinations.
    Section 73.120(c)(7) would align with the corresponding 
requirements under Sec.  37.23(f), and Sec.  73.120(c)(8) would align 
with the corresponding requirements under Sec.  37.31. These 
requirements would encompass the roles and responsibilities for 
licensees, applicants, and if applicable, the contractor/vendors to 
establish, implement, and maintain a system of files and records to 
ensure personal information is not disclosed to unauthorized persons.
    Section 73.120(c)(9) would align with the requirements of Sec.  
37.33. Section 73.120(c)(10) would require licensees, applicants, and 
contractors or vendors to maintain the records that are required by the 
regulations in this section and retain them for a period of 3 years 
after the record is superseded or no longer needed. The record 
retention period of three years would be consistent with Sec.  
37.23(h), contrasting with the five-year retention period under Sec.  
73.56(o). Records maintained in any database(s) would need to be 
available for NRC review, consistent with the requirements found under 
Sec.  73.56(o)(6)(ii).

VI. Specific Requests for Comments

    The NRC is seeking advice and recommendations from the public on 
this proposed rule. We are particularly interested in comments and 
supporting rationale from the public on the following:

Part 26--Fitness for Duty Program

    1. The proposed rule under Sec.  26.603(c) would enable a licensee 
or other entity to implement an FFD program under proposed Sec.  
26.604, ``FFD program requirements for facilities that satisfy the 
Sec.  26.603(c) criterion,'' if the

[[Page 86981]]

licensee or other entity performs a site-specific analysis to 
demonstrate that the facility and its operation satisfy the criterion 
in Sec.  53.860(a)(2).
    Should the NRC consider replacing its proposed Sec.  26.603(c) 
criterion referencing Sec.  53.860(a)(2) with an alternative 
requirement that if the commercial nuclear plant is of the class 
described in Sec.  53.800, ``Facility licensees for self-reliant-
mitigation facilities,'' and either Sec.  53.800(a)(1) or (2) is 
satisfied, then drug and alcohol testing would not be required? This 
proposal would align the Sec.  26.603(c) criterion with that proposed 
in the NRC-licensed operator regulatory framework of part 53. Please 
provide your considerations and rationale for your recommendation.
    Should the NRC also consider making a conforming change to the 
proposed Sec.  73.120 criterion used for the AA program? Please provide 
your considerations and rationale for your recommendation.

Part 26--Technology-Inclusive Approaches to Fatigue Management 
Requirements Applicable to Unit Outages

    In establishing the outage minimum days off requirement of Sec.  
26.205(d)(4), the NRC's objective was to ensure that individuals 
performing the duties described in Sec.  26.4(a)(1) through (a)(4) have 
sufficient periodic long-duration breaks to prevent cumulative fatigue 
from degrading their ability to safely and competently perform their 
duties. In addition to the science of fatigue management, the NRC 
considered several factors in establishing the existing requirements. 
These additional factors were practical and safety considerations 
associated with the management of refueling outages for large LWRs, 
including the following: (1) the typical duration and frequency of 
outages; (2) the availability of contract personnel to perform the 
work; (3) the risk presented by the outage work while the reactor is 
shut down; and (4) the controls applied to the work that may limit the 
potential for latent errors to challenge reactor safety when the 
reactor is returned to power. The details of such considerations may 
differ for new reactor technologies or designs. Such considerations may 
not be relevant for some reactor designs (e.g., reactors capable of on-
line refueling) and there may be additional, more pertinent factors to 
consider for other designs.
    The NRC is seeking stakeholder input on whether alternative fatigue 
management requirements applicable to outages should be adopted to 
support technology-inclusive approaches that would be appropriate to 
support the licensing and regulation of future commercial nuclear 
plants. Please provide your considerations and rationale for your 
recommendation.

Part 26--Draft Regulatory Guidance Approach for Fatigue Management

    In support of this proposed rule, the NRC has issued DG-5078, 
``Fatigue Management for Nuclear Power Plant Personnel at Commercial 
Nuclear Plants Licensed Under 10 CFR part 53.'' This DG describes 
methods the NRC staff considers acceptable for addressing certain 
aspects of FFD programs at commercial nuclear facilities licensed under 
part53.
    The NRC staff also intends to eventually transition this draft 
guide into an update to RG 5.73, ``Fatigue Management for Nuclear Power 
Plant Personnel,'' or the development of a new RG. At this point, NRC 
staff is considering four options for future RG development:
     Option 1: Amend the existing RG. The NRC may develop an 
updated version of RG 5.73 that continues to endorse (with 
clarifications, additions, and exceptions) the guidance contained in 
NEI 06-11, ``Managing Personnel Fatigue at Nuclear Power Reactor 
Sites,'' Revision 1, and incorporates the topics discussed within DG-
5078 as new NRC staff positions in section C of RG 5.73.
     Option 2: Issue a new RG specific to part 53 licensees. 
The NRC may develop an entirely new RG applicable specifically to 
facilities licensed under part 53. This new RG would capture the 
guidance contained in DG-5078 and incorporate existing guidance (e.g., 
selected guidance in RG 5.73 and NEI 06-11) that is considered to be 
technology inclusive in nature. The existing guidance (i.e., RG 5.73) 
would remain in place as the guidance for facilities licensed under 
parts 50 and 52.
     Option 3: Review and potentially endorse new or revised 
industry-developed guidance. The NRC may engage with the industry 
regarding a potential update to industry guidance document NEI 06-11 or 
the development of new, separate industry-developed guidance specific 
to facilities licensed under part 53. The NRC would then review the new 
or revised industry-developed guidance within the NRC's RG process, 
which includes opportunities for public participation. New or revised 
industry-developed guidance could incorporate DG-5078 or propose 
alternatives for the NRC to consider.
     Option 4: Develop a comprehensive revision of the existing 
RG. The NRC may develop a more comprehensive revision of RG 5.73 that 
would explicitly detail all NRC positions reflected in the existing RG 
(including those endorsed positions currently contained in NEI 06-11, 
Revision 1), along with the guidance of DG-5078. Such a revision would 
thereby be a ``stand-alone'' document, without reference to or explicit 
endorsement of separate, industry-developed guidance.
    The NRC is seeking stakeholder input regarding which of the four 
options listed above would be optimal (or whether there are other 
options that the NRC should consider). Please provide your 
considerations and rationale for your recommendation.

Part 53--Overall Organization

    Part 53 is structured as one framework with subparts providing 
technical, licensing, and administrative requirements for the various 
stages of the life cycle of a commercial nuclear plant. The 
organization of part 53 in this manner puts a complete set of 
requirements for each stage of the life cycle in a separate subpart 
with additional subparts for licensing and administrative requirements.
    The NRC is seeking comment on the proposed organization of the 
requirements in part 53 and possible improvements to how specific 
requirements (e.g., examples of which specific sections) could be 
consolidated or otherwise reorganized to make the rule clearer or more 
concise.
    There are numerous references in proposed part 53 to other NRC 
regulations. Examples of such references include those in proposed 
Sec.  53.610 to NRC regulations related to radiation protection (part 
20), FFD (part 26), physical security (part 73), and MC&A (10 CFR part 
74, ``Material Control and Accounting of Special Nuclear Material'') 
for facilities receiving byproduct or SNMs.
    The NRC is seeking comment on whether such references to other 
regulations in various sections in the proposed part 53 provide 
benefits to applicants and licensees, or to other stakeholders seeking 
to understand the regulatory framework under part 53, or whether such 
references could be removed to reduce the length of part 53.

Part 53, Subpart B--Comprehensive Risk Metrics

    The NRC is proposing to require the use of comprehensive risk 
metrics and associated risk performance objectives as one of several 
performance standards in part 53. Comprehensive risk metrics could 
include a risk metric or set of risk metrics that approximate the total 
overall risk from the facility to the

[[Page 86982]]

extent practicable. Associated risk performance objectives are 
preestablished values indicative of the comprehensive risk metrics that 
are used during risk-informed decision-making to gauge plant safety. 
Specifically, comprehensive risk metrics and associated risk 
performance objectives would provide one element of the safety criteria 
for LBEs other than DBAs in the proposed Sec.  53.220. Comprehensive 
risk metrics, in the form of the IEFR and the ILCFR, and associated 
risk performance objectives, in the form of the QHOs of 
5x10-7 per year and 2x10-6 per year, 
respectively, were similarly used in the LMP methodology to ensure that 
other evaluation criteria were conservatively defined and as a tool for 
focusing attention on matters important to managing the risks posed by 
nuclear power plants. The use of such comprehensive risk metrics and 
associated risk performance objectives in an integrated risk-informed 
decision-making process is similar to that used in RG 1.174, ``An 
Approach for Using Probabilistic Risk Assessment in Risk-Informed 
Decisions on Plant-Specific Changes to the Licensing Basis,'' Revision 
3.
    The NRC is seeking comment on the use of comprehensive risk metrics 
and associated risk performance objectives in part 53 as one of several 
performance standards. The IEFR and ILCFR and the QHOs represent 
comprehensive risk metrics and associated risk performance objectives 
that the NRC has used for decades in a variety of capacities. What 
other performance standards could be used to address the comprehensive 
risks posed by proposed commercial nuclear plants? Please provide your 
considerations and rationale for your recommendation.
    If an applicant proposes a novel approach to comprehensive plant 
risk and the NRC approves the approach, should the resulting NRC-
approved comprehensive plant risk metrics and associated risk 
performance objectives be codified or otherwise memorialized over time 
and, if so, how?

Part 53, Subpart B--Defense in Depth

    Proposed Sec.  53.250 would establish requirements based on the 
longstanding NRC philosophy of providing defense in depth to address 
uncertainties concerning the design, operation, and performance of 
commercial nuclear plants during LBEs.
    The NRC is seeking comment on the inclusion of the proposed 
requirements to assess and provide defense in depth. The NRC is also 
seeking comment on whether to include specific provisions in Sec.  
53.250 and subpart B to more explicitly address the possible role of 
inherent characteristics of some SSCs in preventing or mitigating 
unplanned events. The proposed Sec.  53.250 is worded to preclude 
relying on a single engineered design feature to address the range of 
LBEs other than DBAs, which could possibly allow crediting inherent 
characteristics without further lines of defense. How could possible 
inherent characteristics of SSCs be considered in the proposed 
requirements in Sec.  53.250 or in any alternative requirements for 
defense in depth provided in response to this item? Please provide your 
considerations and rationale for your recommendation.

Part 53, Subpart C--Probabilistic Risk Assessment

    Current consensus PRA standards provide processes for appropriately 
defining the scope of a PRA and determining applicability of supporting 
requirements to suit the specific needs of a given applicant under 
proposed part 53. In addition to assessing other aspects of PRA 
acceptability such as PRA peer reviews, NRC determinations of the 
acceptability of such PRAs would assess the appropriateness of the 
applicant-defined scope as part of determining the applicability of a 
consensus PRA standard supporting an application. This approach is 
consistent with the current state of practice and offers appropriate 
flexibility for PRAs to be developed and assessed based on the 
application they are used to support, which includes consideration of 
how PRA results and insights are relied upon, together with factors 
such as safety margin, simplicity of design, and treatment of 
uncertainty.
    The NRC is seeking comment on what additional guidance, if any, is 
needed regarding PRA acceptability for Part 53 applicants and 
licensees.

Part 53, Subparts C and D--Earthquake Engineering

    Proposed Sec.  53.480 would establish requirements related to 
seismic design considerations. This proposed section is intended to 
provide a clear connection between siting activities and seismic design 
activities and to support various approaches to presenting seismic 
hazards and addressing those hazards in designs. The proposed 
requirements are intended to provide sufficient flexibility to allow 
approaches like those currently in parts 50 and 100 or approaches that 
might be endorsed by the NRC in the future that could incorporate more 
risk insights from PRAs.
    The NRC is seeking comment on whether the proposed requirements for 
earthquake engineering provide appropriate flexibility in addressing 
seismic risks while also ensuring that the regulations continue to 
adequately address seismic hazards. Please provide your considerations 
and rationale for your recommendation.

Part 53, Subpart E--Construction and Manufacturing

    1. Proposed Sec.  53.610(b)(1)(iii) would require procedures that 
describe how construction will be controlled so as not to impact other 
features important to the design (e.g., dewatering, slope stability, 
backfill, compaction, and seepage).
    The NRC is seeking comment on whether such specific requirements 
are useful or whether these requirements could be met through other 
requirements proposed in part 53 or already present in other relevant 
regulations (e.g., quality assurance requirements in appendix B to part 
50).

Part 53, Subparts E and H--Manufacturing Licenses

    1. The proposed requirements governing manufacturing are set forth 
in subpart E, and the proposed requirements governing the licensing 
processes are contained in subpart H. Some of the proposed 
requirements, including provisions related to the loading of 
unirradiated fuel into a manufactured reactor, are intended to cover a 
factory-fabrication model that has been suggested for some micro-
reactor designs. However, as written, the proposed provisions are not 
limited to any size or type of reactor.
    The NRC is seeking comment on whether the proposed regulations are 
sufficient to govern various scenarios for the possible manufacturing 
and deployment of manufactured reactors.
    If a comment indicates that the proposed regulations are not 
sufficient, please describe the reasons why, including, if applicable, 
any plausible scenario for which the commenter believes the proposed 
regulations are not sufficient.
    2. The proposed regulations in subpart H allow holders of or 
applicants for a COL to reference an ML but do not include such a 
provision for the holder of or applicant for a CP or OL. This proposed 
change from the current relationship between subparts in part 52 and 
the part 50 licensing process was made to simplify the provisions in 
the proposed part 53 for licensing and deploying manufactured reactors.
    The NRC seeks comment on whether part 53 should include provisions 
for an applicant for or a holder of a CP or an OL to reference an ML 
and, if so, how this should be done.

[[Page 86983]]

    3. Proposed Sec.  53.1295 states that the holder of an ML could not 
begin manufacture of a manufactured reactor less than 6 months before 
the expiration of the license. This limitation is similar to the 
current restriction in Sec.  52.177, which states that the manufacture 
of a reactor cannot begin less than 3 years before the expiration of 
the license. The restriction was revised from 3 years in part 52 to 6 
months in the proposed part 53 in recognition of the likely use of MLs 
for a factory-fabrication model for micro-reactors.
    The NRC seeks comment on whether it is necessary or appropriate to 
revise the 3-year restriction in part 52 on when manufacturing 
activities could begin in relation to license expiration and, if so, 
what that restriction should be.
    4. Proposed Sec.  53.1288 provides the finality provisions for MLs 
and includes, as does existing Sec.  52.171, limitations on the NRC's 
imposition of new requirements on either the design or the requirements 
for the manufacture of a manufactured reactor. No MLs have been issued 
under part 52 and there is no practical experience with the proposed 
finality sections. While the implications of the finality provisions 
related to the design of a manufactured reactor can reasonably be 
inferred from experience with DCs and COLs, there is no experience or 
available guidance regarding finality for ``requirements for the 
manufacture of the manufactured reactor.''
    The NRC is seeking comment on the proposed finality provisions for 
MLs and specifically if and how finality for manufacturing processes 
might be requested and used.
    5. The NRC is seeking comment on the proposed regulations for the 
loading of fresh (unirradiated) fuel into a manufactured reactor for 
subsequent transport to a site for which the Commission has issued a 
COL that authorizes construction and operation of a commercial nuclear 
plant using the manufactured reactor. The proposed regulation includes 
provisions for loading of fuel into manufactured reactors at a 
manufacturing facility prior to transporting the fueled reactor to its 
deployment site, as suggested by some stakeholders. The NRC has 
historically viewed reactor operation as including fuel load, and 
existing NRC regulations reflect this view. While the Act authorizes 
the NRC to issue licenses to manufacture production or utilization 
facilities, it does not contain specific provisions on fueling or 
operating facilities licensed under an ML, and existing ML regulations 
under part 52 do not include provisions for fuel load.
    The proposed rule addresses this matter by allowing an applicant to 
combine an ML with a part 70 license, which would authorize possession 
of a manufactured reactor in which the licensee has loaded unirradiated 
fuel provided at least two independent criticality prevention 
mechanisms are in place, each of which is sufficient to prevent 
criticality assuming optimum neutron moderation and neutron reflection 
conditions. This requirement would limit the possibility of creating 
fission products and allow the control of SNM, so that the loading of 
the fuel into a manufactured reactor could be governed primarily via a 
part 70 license and associated regulations (including those in subpart 
H of part 70).
    A specific topic on which the NRC is seeking comment is on the 
potential benefits of and issues with including the requirements of 
subpart H of part 70 within the proposed regulations for loading fuel 
into manufactured reactors at the manufacturing facility. For example, 
should the NRC include a threshold for including the requirements of 
subpart H of part 70 and, if so, what factors and decision criteria 
should be considered in such a threshold? If a comment indicates that 
the proposed regulations are not sufficient, please describe the 
reasons why, including the plausible scenarios for which the proposed 
regulations would not work or could be made to work better.
    6. Section 170, ``Indemnification and Limitation of Liability,'' of 
the Act states that each license under section 103 shall have as a 
condition of the license a requirement that the licensee have and 
maintain financial protection of such type and in such amounts as the 
NRC shall require.
    The NRC is seeking comment on whether the proposed regulations 
should include amounts of required financial protections for MLs for 
fueled manufactured reactors, and, if so, what would be appropriate 
amounts of required financial protection.
    7. Some stakeholders have suggested that a fueled manufactured 
reactor with appropriate protections against criticality should not be 
categorized as a utilization facility under NRC regulations or Section 
11cc. of the Act.
    The NRC is seeking comment on possible approaches where the NRC 
could find that a fueled manufactured reactor would not be a 
utilization facility, the basis for such a finding, and the potential 
benefits of and potential issues with such a finding.
    8. Proposed requirement Sec.  53.620(d)(2)(i) would require a 
security program, including a physical security plan, for any ML 
authorizing possession of a manufactured reactor into which fuel has 
been loaded at the manufacturing facility. Currently, requirements in 
Sec.  73.67(c)(1) only require that a physical security plan be 
submitted for those licensees who possess, use, transport, or deliver 
to a carrier for transport SNM of moderate strategic significance, or 
10 kg or more of SNM of low strategic significance.
    The NRC is seeking comment on whether the proposed requirement: (1) 
should be specific to the facility type (i.e., manufacturing facility) 
or be specific to the category of material being used at the facility; 
(2) should apply to all manufacturing plants, including those at which 
licensees may only possess SNM of low strategic significance (i.e., 
category III), or only those facilities for which an applicant must 
submit a physical security plan per Sec.  73.67(c)(1); or (3) should 
include more specific requirements on the supplemental security 
measures that may be needed for licensees possessing SNM of moderate 
strategic significance (i.e., category II)?
    9. Proposed requirement Sec.  53.620(d)(2)(i) would require a 
cybersecurity program. The proposed general cybersecurity performance 
requirements would be to provide reasonable assurance that a 
cyberattack could not adversely impact the functions performed by 
digital assets used by the licensee for implementing the physical 
security, radiation monitoring, and criticality requirements.
    The NRC is seeking comment on the following: (1) to what extent 
stakeholders envision physical security controls, radiation monitoring, 
and criticality controls at a manufacturing facility being digital; (2) 
to what extent should the ML holder be required to protect digital 
computer and communications systems that impact safety and security 
functions from a cyberattack at a manufacturing facility authorized to 
load fuel; and (3) whether the rule provides sufficient clarity on the 
cybersecurity measures needed for license issuance or if additional 
detail should be included either in the rule or in guidance?
    10. Proposed requirement Sec.  53.620(d)(2)(i)(B) would require 
that the physical security program be designed to prevent unintended 
and uncontrolled criticality events. This would include criticality 
events that are initiated maliciously.
    The NRC is seeking comment on whether the ML holder should be 
required to design its security program to protect against radiological 
sabotage

[[Page 86984]]

(i.e., an unintended criticality event leading to unacceptable 
radiological consequences), in addition to theft and diversion. For 
example, should the NRC establish security requirements to prevent an 
adversary, including an insider, from tampering with the reactor at a 
manufacturing facility or during transport in such a way as to cause an 
inadvertent criticality event? If so, should the NRC consider factors 
such as the category of fuel and the number of reactors at a factory 
that can simultaneously be loaded with fuel in establishing the 
security requirements?
    11. Proposed requirement Sec.  53.620(d)(2)(i) would require an ML 
holder to meet the performance objectives in Sec.  73.67. Requirements 
Sec.  73.67(e) and Sec.  73.67(g) include provisions for security of 
category II and category III quantities of SNM, respectively, during 
transportation.
    The NRC is seeking comment on the extent to which the ML should 
require ASMs (i.e., security measures above those required by Sec.  
73.67(e) and Sec.  73.67(g)) for transportation of a fueled reactor to 
its place of operation. What should those measures be?
    12. Proposed requirement Sec.  53.620(d)(2)(i) would require an ML 
holder to meet the performance objectives of Sec.  73.67. For licensees 
utilizing a category II quantity of SNM, the requirement in Sec.  
73.67(d)(4) would have the ML holder conduct a screening to confirm the 
identity of an individual prior to granting unescorted access to the 
controlled access area where the material is used or stored. The 
purpose of this requirement is to both confirm the identity of the 
individual and support a determination that the individual is 
trustworthy and reliable.
    The NRC is seeking comment on whether the ML requirements should 
include ASMs (i.e., measures beyond those required by Sec.  
73.67(d)(4)) in order to provide reasonable assurance of identity 
confirmation and trustworthiness and reliability.
    13. The NRC is seeking comment on whether provisions regulating the 
testing of fueled manufactured reactors in the manufacturing facility 
should be included in part 53 and, if so, what would be practical for 
the holder of an ML while also providing adequate protection of public 
health and safety. One possibility could be COLs that would be issued 
to the holders of an ML to cover low power (e.g., <5% rated thermal 
power) nuclear physics testing of fueled manufactured reactors within 
the manufacturing facility prior to the manufactured reactors being 
transported to and incorporated into a commercial nuclear plant for the 
purpose of energy production. The NRC recognizes configuration changes 
are needed to perform nuclear physics testing and is seeking comment on 
what requirements should apply to the manufactured reactors and the 
manufacturing facility during such testing (e.g., limiting power 
levels). If a comment indicates that the regulations should address 
limited operations at manufacturing facilities, please describe the 
likely scenarios that would need to be addressed and suggest what would 
be appropriate requirements for such scenarios.
    While an ML holder could accomplish nuclear physics testing by 
applying for a COL under the proposed subpart H of part 53, 
stakeholders have indicated that many of the requirements would likely 
be unnecessary, given the reduced risk profile posed by such 
activities. Therefore, the NRC is seeking comment on what requirements 
in subpart H of part 53 should apply to applicants for a COL who would 
perform testing of fueled manufactured reactors at the manufacturing 
plant. Examples of proposed requirements that might be relaxed or 
modified for applications for low power testing at manufacturing plants 
include those related to selection of LBEs to reflect limited inventory 
of radionuclides and decay heat, aircraft impact assessments, and 
earthquake engineering.
    Additionally, the NRC is seeking comment on whether several other 
requirements in part 53 could be modified for applications for a low 
power testing COL at a manufacturing facility. For example, the NRC is 
seeking comment on how portions of the ML facility used to support 
testing should fall within the requirements for construction activities 
under Sec.  53.610; whether Sec. Sec.  53.710 and 53.715 (SSC 
configuration control) must be implemented to ensure portions of the ML 
facility relied on to limit potential radiological consequences from 
LBEs are available to perform their safety functions; and whether the 
requirements of Sec.  53.730 could be modified to reflect the 
conditions of low power physics testing. If a comment indicates that 
some design and analysis requirements and related application 
requirements in subpart H of the proposed part 53 are not needed for 
the testing of fueled manufactured reactors, please provide a rationale 
supporting your comment and, if applicable, what alternate requirements 
would be appropriate.
    Moreover, the licensing mechanism for the facility could present 
unique challenges. One option could be to issue a low power testing COL 
for each fueled manufactured reactor to be tested. This would comport 
with the agency's practice of issuing one license per reactor but could 
prove prohibitive from a cost standpoint and may provide very little 
safety benefit if all manufactured reactors are the same. 
Alternatively, one low power testing COL could be issued for the 
portions of the ML facility used to test the fueled manufactured 
reactors and allow multiple fueled manufactured reactors to be 
completed and tested over the course of the ML. Under this approach, 
any ITAAC related to testing of the fueled manufactured reactors would 
need to be closed after they were manufactured but prior to testing, 
and the NRC would issue a notice of intended operation and provide the 
public an opportunity to request a hearing on whether each fueled 
manufactured reactor as constructed complies, or on completion will 
comply, with the acceptance criteria of the license. The NRC is seeking 
comment on the potential benefits and issues with having a COL for each 
fueled manufactured reactor to be tested versus having a COL cover the 
testing of multiple fueled manufactured reactors. If a comment 
indicates a preference for a particular approach, please provide a 
rationale supporting the comment and describe the specific scenarios 
that the regulations need to address.

Part 53, Subpart F--Staffing and Generally Licensed Reactor Operators

    Under the Act Sections 106 and 107, the NRC is proposing to group 
commercial reactors into classes upon the basis of the similarity of 
operating and technical characteristics of the facilities, and then to 
prescribe uniform conditions for licensing individuals as operators of 
any of the various classes; determine the qualifications of such 
individuals; and, for certain classes of commercial reactors, issue 
general licenses (i.e., licenses for which no application is needed) to 
such individuals allowing the individuals to operate the commercial 
reactor.
    1. Categories of Individuals Who May Manipulate Facility Controls: 
The NRC is proposing requirements that would allow the manipulation of 
the controls of certain facilities by GLROs in lieu of specifically 
licensed reactor operators and senior reactor operators. Reactor 
operators and senior reactor operators are the only categories of 
individuals currently allowed to be licensed to manipulate the controls 
of utilization facilities under part 55.
    The NRC is interested in public perspectives on this proposed 
addition of the GLRO category, particularly in light of new reactor 
technologies and concepts of operations.

[[Page 86985]]

    2. Criteria for GLRO Staffing: The NRC is proposing criteria under 
which facilities would be staffed by GLROs in lieu of specifically 
licensed reactor operators and senior reactor operators. These criteria 
establish a new class of self-reliant-mitigation facilities, as defined 
in part 53, for which distinct GLRO licensing and staffing requirements 
would apply.
    The NRC is soliciting public feedback regarding whether these 
proposed criteria are appropriate and what, if any, alternative 
criteria should be considered. Please provide your considerations and 
rationale for your answer.
    3. Medical Requirements for GLROs: Based on the proposed criteria 
that a self-reliant-mitigation facility, as defined in part 53, must 
meet, the NRC is proposing not to subject GLROs to requirements for 
medical fitness and medical examination. This is in contrast with the 
proposed requirements associated with specifically licensed reactor 
operators and senior reactor operators, as well as the existing 
requirements for reactor operators and senior reactor operators under 
part 55.
    The NRC is soliciting public feedback regarding whether GLROs 
should be subject to medical fitness and/or medical examination 
requirements like reactor operators and senior reactor operators. 
Please provide your considerations and rationale for your answer.
    4. Onshift Engineering Expertise: The NRC is proposing to require 
that engineering expertise be accounted for within facility staffing 
plans. This proposed requirement would be in lieu of the traditional 
position of the Shift Technical Advisor. The NRC is further proposing 
that individuals providing such engineering expertise would need, among 
other things, to possess either a qualifying 4-year degree or licensure 
as a Professional Engineer.
    The NRC is interested in feedback from the public regarding the 
appropriateness of this requirement, including any alternatives that 
should be considered. Please provide your considerations and rationale 
for your answer.
    5. Use of Simulation Facilities as HFE Testbeds: The NRC is 
proposing to establish regulations pertaining to the use of simulation 
facilities within the context of the licensing programs both for 
specifically licensed reactor operators and senior reactor operators as 
well as for GLROs. However, these regulations, as currently proposed, 
do not address the use of simulation facilities within the context of 
serving as testbeds for HFE-related analyses and assessments. Rather, 
the NRC currently envisions that the use of simulation facilities as 
HFE testbeds is more appropriately addressed via guidance documents.
    The NRC is soliciting public feedback regarding whether simulation 
facility requirements should also address the use of simulation 
facilities as HFE testbeds. Please provide your considerations and 
rationale for your answer.

Part 53, Subpart F--Emergency Preparedness and Security Programs

    1. The proposed framework for part 53 would incorporate the changes 
to NRC regulations from the final rulemaking on ``Emergency 
Preparedness for Small Modular Reactors and Other New Technologies'' 
(the EP for SMR/ONT rule) by including references to Sec.  50.160, 
``Emergency preparedness for small modular reactors, non-light-water 
reactors, and non-power production or utilization facilities,'' and by 
making conforming changes within Sec.  50.160. The proposed framework 
for part 53 would also introduce a graded approach to physical 
protection requirements that includes the criterion in Sec.  
53.860(a)(2)(i) to establish a class of licensees that would not be 
required to protect against the design-basis threat (DBT) of 
radiological sabotage. The NRC is soliciting public comment relating to 
these topics, which could include ways that graded approaches for both 
emergency preparedness and security programs might be assessed and 
considered during the licensing process.
    The NRC is seeking comment on the sufficiency and clarity of 
requirements in proposed part 53 related to the assessments needed to 
support graded emergency planning and security. If a comment indicates 
that there is an issue with the sufficiency or clarity of the proposed 
regulations, please describe the reasons why, including, if applicable, 
any scenario for which the proposed regulations are not sufficient and 
possible ways to clarify the requirements. The NRC is specifically 
seeking comment on possible challenges arising from the interactions 
between the proposed regulations and related assessments for grading 
the requirements for emergency planning and security.
    2. The NRC is preparing various guidance documents to support this 
rulemaking and other ongoing or recently completed rulemakings related 
to emergency preparedness and security. DG-5076, ``Guidance for 
Technology-Inclusive Requirements for Physical Protection of Licensed 
Activities at Commercial Nuclear Plants,'' has been issued along with 
this proposed rulemaking and public comments are requested via this 
notice on that draft guidance. The NRC is also planning to issue a 
draft revision of RG 1.242, ``Performance-Based Emergency Preparedness 
for Small Modular Reactors, Non-Light-Water Reactors, and Non-Power 
Production or Utilization Facilities,'' for public comment. The planned 
revision to RG 1.242 would add guidance for part 53 applicants and 
licensees.
    In the staff requirements memorandum to SECY-23-0021, the 
Commission directed the NRC staff to address the consideration of 
security-related events for an advanced reactor that addresses security 
through design and engineered safety features when it harmonizes this 
rulemaking with the EP for SMR/ONT rule. In the EP for SMR/ONT rule, 
the NRC established an alternative performance-based and risk-informed 
approach for emergency planning, including determining the need for and 
size of an emergency planning zone (EPZ) to support predetermined, 
prompt protective actions. The NRC has incorporated the relevant rule 
language from the EP for SMR/ONT rule into this proposed rule and is 
seeking stakeholder feedback as to whether additional rule language 
changes or additional guidance would be beneficial.
    In light of the Commission direction and the above considerations, 
the NRC is assessing how best to address the treatment of security-
related events in emergency planning, including in the determination of 
EPZ size, for reactors licensed under part 53. Part 53 is introducing 
an alternative approach to meeting security regulations that should be 
taken into consideration under Sec.  50.160. Stakeholders are 
encouraged to take a holistic view of the various activities and 
opportunities to provide comments on this rulemaking and related 
guidance supporting this rulemaking (e.g., DG-5076 on physical 
protection requirements, future revisions to RG 1.242). In developing 
comments, the NRC urges stakeholders to consider various scenarios that 
might arise when implementing graded approaches for security and 
emergency planning for various reactor designs. Scenarios could include 
the following:
     the potential consequences from security events up to and 
including the DBT of radiological sabotage are bounded by unlikely and 
very unlikely event sequences such that security events do not need 
separate analyses in the EPZ size determination;

[[Page 86986]]

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

Part 53, Subpart F--Integrity Assessment Program Requirements

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

Part 53, Subpart G--Decommissioning

    1. On March 3, 2022, the NRC published the proposed rule entitled 
``Regulatory Improvements for Production and Utilization Facilities 
Transitioning to Decommissioning'' (87 FR 12254). This rulemaking would 
amend the NRC's current regulations to provide an appropriate 
regulatory framework for nuclear power reactors transitioning from 
operations to decommissioning. The rulemaking would address lessons 
learned from licensees that have completed or are currently in the 
decommissioning process. The NRC staff sent a draft final rule to the 
Commission for its consideration on January 31, 2024, in SECY-24-0011, 
``Final Rule: Regulatory Improvements for Production and Utilization 
Facilities Transitioning to Decommissioning (3150-AJ59; NRC-2015-
0070).''
    What aspects of this draft final rule, if any, should be 
incorporated in a part 53 final rule and why?
    2. Proposed Sec.  53.1060(b) in subpart G would require that, ``No 
later than 30 days after the Commission publishes notice in the Federal 
Register under Sec.  53.1452(a), the licensee must submit a report 
containing a certification that financial assurance for decommissioning 
is being provided in an amount specified in the licensee's most recent 
updated certification, including a copy of the financial instrument 
obtained to satisfy Sec.  53.1040.'' This is similar to the current 
requirement in Sec.  50.75(e)(3) for part 52 COL holders. The NRC is 
seeking comment on whether commercial nuclear plant COL holders under 
part 53 should have the same requirement as COL holders under part 52 
to demonstrate that they have financial assurance in place no later 
than 30 days after the Commission issues the notice of intended 
operation under Sec.  53.1452. Please provide your considerations and 
rationale for your answer.

Part 53, Subpart H--Licenses To Construct and Operate Commercial 
Nuclear Plants of Identical Design at Multiple Sites

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

[[Page 86987]]

safety of a referenced design certification or manufactured reactor.
    Given this synopsis of how the requirements in proposed Sec.  
53.1470 would be implemented as currently written, the NRC is seeking 
comment on whether there are opportunities to allow added flexibility 
for applicants under these provisions. This could include consideration 
of whether applications for which the ``common design'' is not 
completely identical could be evaluated under this provision and, if 
so, what the process would be for determining the appropriateness of a 
common review. In addition, the NRC is interested in feedback about the 
pros and cons of requiring that applications under these proposed 
provisions be submitted at the same time versus allowing them to be 
submitted on a staggered basis.

Part 53, Subparts H and I--Probabilistic Risk Assessment Information

    Proposed Sec.  53.1239(a)(18) in subpart H and the related 
references to this proposed requirement for the holders of OLs and COLs 
would require a description of the PRA required by Sec.  53.450(a), and 
its results to be included in FSARs. However, guidance documents may 
further clarify the division of PRA-related information needed to be in 
the FSAR, in other possible licensing basis documents, and controlled 
as plant records subject to inspections and audits. For example, a 
possible approach could be to include a summary of the PRA results in 
the FSAR and control that information under Sec.  53.1545 and create a 
separate document related to the broader PRA analyses and related 
processes as a program document under Sec.  53.1560. The program 
document would provide more detail than the summaries in the FSAR but 
still be a much-condensed source of information in comparison to the 
documentation of the PRA. This possible approach would reflect the role 
of the PRA in the licensing process under part 53 and in maintaining 
margins to the safety and evaluation criteria in subparts B and C but 
may allow a more appropriate evaluation process to address the 
particulars and complexities of the PRA-related documents.
    The NRC is seeking comment on the appropriate placement of PRA-
related information among various licensing basis documents and plant 
records. In addition to the placement of PRA-related information, the 
NRC is seeking comment on the appropriate control of that information 
and on the routine submittal of updates to the NRC. Please provide your 
considerations and rationale for your answer.

Part 53, Subparts H and I--Changes to Manufacturing Licenses

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

    \15\ Section 70.23(a)(5).
---------------------------------------------------------------------------

    If a part 70 financial qualification standard were to apply to a 
category of merchant new reactor applicants, should it also apply to 
pre-construction license transfer applications for these reactors? Why 
or why not?
    Is there another standard the agency should consider for financial 
qualification of merchant new reactor applicants? Commenters are 
encouraged to provide specific suggestions and the basis for those 
suggestions.

Part 73, Section 73.100--Physical Security

    The proposed Sec.  73.100 would identify the proposed performance-
based physical security requirements with which future commercial power 
reactor applicants or licensees' physical protection programs would 
need to demonstrate compliance, without prescribing the specific 
methods that must be used to satisfy them. Applicants and licensees 
would have increased flexibility regarding the modern technologies and 
methods that they could use. Implementing guidance in DG-5076 (proposed 
RG 5.97), ``Guidance for Technology Inclusive Requirements for Physical 
Protection of Licensed Activities at Commercial Nuclear Plants,'' would 
be available to assist applicants and licensees. For example, DG-5076 
provides detailed guidance, including performance standard 
recommendations, on the probability of detection and alternative 
sources of power for exterior intrusion detection systems (subsection 
4.1.1.1.A), interior intrusion detection (subsection 4.1.1.1.B), 
intrusion assessment (subsection 4.1.1.2.A), security response/
neutralization subsection (4.1.1.4.A), security communication 
(subsection 4.1.1.3.A), and security delay (subsection 4.1.1.4.C).
    Does the NRC's proposed approach in Sec.  73.100 provide a 
sufficient level of detail to be readily understood and easily applied 
to the licensing and oversight of new and advanced power reactors, or 
should the NRC consider moving some objective and measurable security 
performance standard recommendations from the draft implementing 
guidance in DG-5076 into proposed Sec.  73.100? If so, which objective 
and measurable security performance standard recommendations should be 
moved from DG-5076 to Sec.  73.100? Please provide the basis for your 
response.

[[Page 86988]]

Part 73, Section 73.110--Cybersecurity

    The proposed Sec.  73.110 would require licensees to demonstrate 
protection against cyberattacks in a manner that is commensurate with 
the potential consequences from those attacks, without prescribing the 
specific methods that must be used to demonstrate protection. Under 
proposed Sec.  73.110(a), licensees would need to ensure that digital 
computer and communications systems are adequately protected against a 
potential cyberattack that would, for example, result in adverse 
impacts to the physical security digital assets used by the licensee to 
prevent unauthorized removal of material per Sec.  53.860(a). 
Protecting against such a potential cyberattack would involve requiring 
cybersecurity for SNM at a commercial nuclear reactor licensed under 
part 53. Applicants and licensees would have increased flexibility 
regarding the modern technologies and methods that they could use for 
protecting against such a potential cyberattack. Detailed implementing 
guidance in DG-5075 (proposed RG 5.96), ``Establishing Cybersecurity 
Programs for Commercial Nuclear Plants licensed under 10 CFR part 53,'' 
would be available to assist applicants and licensees. For example, DG-
5075 provides guidance on the implementation of security by design 
features (e.g., facility design) for negating the potential 
consequences from such a potential cyberattack.
    If a cyberattack were to compromise the availability, integrity, or 
confidentiality of data or systems associated with security systems/
measures for the protection of SNM at a commercial nuclear reactor 
licensed under part 53, do the potential consequences warrant requiring 
cybersecurity for such material? Please provide the basis for your 
response including a detailed explanation of challenges, if any, posed 
by requiring cybersecurity for SNM at a commercial nuclear reactor 
licensed under part 53.
Recent Legislation
    On July 9, 2024, the President signed into law the Accelerating 
Deployment of Versatile, Advanced Nuclear for Clean Energy Act of 2024, 
also referred to as the ADVANCE Act. Section 203, ``Licensing 
Considerations Relating to Use of Nuclear Energy for Nonelectric 
Applications,'' and Section 208, ``Regulatory Requirements for Micro-
Reactors,'' of the ADVANCE Act specifically mention the technology-
inclusive regulatory framework to be established under section 
103(a)(4) of NEIMA as a potential vehicle to be considered for the 
report to Congress required under section 203 and a potential vehicle 
to implement strategies and guidance for the licensing and regulation 
of micro-reactors required under section 208. This proposed rulemaking 
is, in part, how the NRC is implementing section 103(a)(4) of NEIMA.
    The NRC is seeking comment on how part 53 could be revised to 
better enable its potential use to implement the ADVANCE Act. 
Specifically, Section 208 of the ADVANCE Act requires the NRC to 
develop and implement ``risk-informed and performance-based strategies 
and guidance'' in several areas for the licensing and regulation of 
micro-reactors, including with respect to ``licensing mobile 
deployment.'' The ADVANCE Act requires the NRC to consider ``the unique 
characteristics of micro-reactors,'' including physical size, design 
simplicity, and source term; opportunities to incorporate specific 
improvements related to streamlining the review process; and other 
policy and licensing issues. With regard to implementation, the ADVANCE 
Act provides the NRC with three options. The NRC may implement the 
developed strategies and guidance, as appropriate, via (1) the existing 
regulatory framework, (2) the Part 53 rulemaking, or (3) a pending or 
new rulemaking. Given the language included in Section 208, the NRC is 
seeking comment on how part 53 could be revised to better address the 
ADVANCE Act's requirements related to strategies and guidance for 
micro-reactors.

VII. Section-by-Section Analysis

    The following paragraphs describe the specific changes proposed by 
this rulemaking.

Sec.  1.43 Office of Nuclear Reactor Regulation

    This proposed rule would revise Sec.  1.43(a)(2) to extend the 
authority of the Office of Nuclear Reactor Regulation to regulate 
source, byproduct, and SNM at facilities licensed under part 53.

Sec.  2.1 Scope

    This proposed rule would revise Sec.  2.1(e) to apply to standard 
design approvals under part 53.

Sec.  2.4 Definitions

    This proposed rule would revise Sec.  2.4 to update the definition 
of ``Contested proceeding'' to include NRC enforcement actions against 
applicants for a standard DC under part 53. It would also update the 
definition of ``Facility'' to encompass utilization facilities as 
defined in Sec.  53.020 (there are no production facilities under part 
53).

Sec.  2.100 Scope of Subpart

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

Sec.  2.101 Filing of Application

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

Sec.  2.104 Notice of Hearing

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

Sec.  2.105 Notice of Proposed Action

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

Sec.  2.106 Notice of Issuance

    This proposed rule would revise Sec.  2.106 to extend the issuance 
notice requirement to applications concerning facilities covered under 
part 53 through updated references in paragraphs (a)(2) and (3), and 
(b)(2).

Sec.  2.109 Effect of Timely Renewal Application

    This proposed rule would revise Sec.  2.109 to add references to 
part 53 in paragraphs (b), (c), and (d) regarding the timing of license 
renewal applications.

Sec.  2.110 Filing and Administrative Action on Submittals for Standard 
Design Approval or Early Review of Site Suitability Issues

    This proposed rule would revise Sec.  2.110 to include references 
to part 53 in paragraphs (a)(1) and (b).

Sec.  2.202 Orders

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

[[Page 86989]]

Sec.  2.309 Hearing Requests, Petitions To Intervene, Requirements for 
Standing, and Contentions

    This proposed rule would revise Sec.  2.309 to include references 
to part 53 in paragraphs (a), (f)(1)(i), (f)(1)(vi) and (vii), (g), 
(h)(2), (i)(2), and (j) regarding a request for hearing under Sec.  
53.1452.

Sec.  2.310 Selection of Hearing Procedures

    This proposed rule would revise Sec.  2.310 by revising paragraph 
(a), the introductory text for paragraph (h), and paragraphs (i) and 
(j) to incorporate references to part 53 regarding hearing procedures.

Sec.  2.329 Prehearing Conference

    This proposed rule would revise Sec.  2.329(a) to extend the timing 
requirements for prehearing conferences involving CPs and licenses 
under part 53.

Sec.  2.339 Expedited Decision-Making Procedure

    This proposed rule would revise Sec.  2.339(d) to include 
references to part 53 regarding expedited decision-making procedures.

Sec.  2.340 Initial Decision in Certain Contested Proceedings; 
Immediate Effectiveness of Initial Decisions; Issuance of 
Authorizations, Permits and Licenses

    This proposed rule would revise Sec.  2.340 regarding initial 
decisions of a presiding officer in certain contested proceedings, the 
effective date of those decisions, and the issuance of authorizations, 
permits, and licenses, by incorporating references to part 53 in 
paragraphs (b), (c), (d), (f), (i), and (j).

Sec.  2.341 Review of Decisions and Actions of a Presiding Officer

    This proposed rule would revise Sec.  2.341(a)(1) to include an 
updated reference to part 53 regarding the allowance of a period of 
interim operation.

Sec.  2.400 Scope of Subpart

    This proposed rule would revise Sec.  2.400 to extend the scope of 
subpart D of part 2 to include part 53 applicants for licenses to 
construct or operate nuclear power reactors of identical design at 
multiple sites.

Sec.  2.401 Notice of Hearing on Construction Permit or Combined 
License Applications Pursuant to Appendix N of 10 CFR Parts 50, 52, or 
53

    This proposed rule would revise the section heading and Sec.  2.401 
to extend the hearing notice requirement to applications concerning 
facilities covered under part 53.

Sec.  2.402 Separate Hearings on Separate Issues; Consolidation of 
Proceedings

    This proposed rule would revise Sec.  2.402(a) to apply provisions 
regarding separate hearings and the consolidation of proceedings to 
part 53 applicants.

Sec.  2.403 Notice of Proposed Action on Applications for Operating 
Licenses Pursuant To Appendix N of 10 CFR Part 50

    This proposed rule would revise Sec.  2.403 to require the 
Commission to publish a notification of proposed action in the Federal 
Register after applications under part 53 are docketed.

Sec.  2.404 Hearings on Applications for Operating Licenses Pursuant to 
Appendix N of 10 CFR Part 50

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

Sec.  2.405 Initial Decisions in Consolidated Hearings

    This proposed rule would revise Sec.  2.405 to be applicable to 
CPs, full-power OLs, and COLs under part 53.

Sec.  2.406 Finality of Decisions on Separate Issues

    This proposed rule would revise Sec.  2.406 to be applicable to 
proceedings conducted pursuant to part 53.

Sec.  2.500 Scope of Subpart

    This proposed rule would revise Sec.  2.500 to extend the 
provisions of subpart E of part 2 to include applications for a license 
to manufacture nuclear power reactors under part 53.

Sec.  2.501 Notice of Hearing on Application Under Subpart F of 10 CFR 
Part 52 or 53 for a License To Manufacture Nuclear Power Reactors

    This proposed rule would revise the section heading and Sec.  
2.501(a) by extending its provisions to applications for a license to 
manufacture nuclear power reactors under part 53.

Sec.  2.643 Acceptance and Docketing of Application for Limited Work 
Authorization

    This proposed rule would revise Sec.  2.643(b) regarding the 
acceptance and docketing of an application for a CP for a utilization 
facility of the type specified in part 53.

Sec.  2.645 Notice of Hearing

    This proposed rule would revise Sec.  2.645(a) to incorporate a 
reference to part 53.

Sec.  2.649 Partial Decisions on Limited Work Authorization

    This proposed rule would revise Sec.  2.649 to extend its 
provisions to LWAs issued under part 53.

Sec.  2.800 Scope and Applicability

    This proposed rule would revise Sec.  2.800 by revising paragraphs 
(c) and (d) to incorporate references to part 53 regarding the scope 
and applicability of the rulemaking procedures contained in this 
subpart.

Sec.  2.801 Initiation of Rulemaking

    This proposed rule would revise Sec.  2.801 to include a reference 
to part 53.

Sec.  2.813 Written Communications

    This proposed rule would revise Sec.  2.813(a) to apply general 
requirements for correspondence with the Commission to communications 
concerning part 53, in addition to parts 50, 52, and 100.

Sec.  2.1103 Scope of Subpart K

    This proposed rule would revise the first sentence of Sec.  2.1103 
to extend the provisions of subpart K of part 2 to licenses under part 
53 to expand the spent fuel capacity at the site of a civilian nuclear 
power plant.

Sec.  2.1202 Authority and Role of NRC Staff

    This proposed rule would amend Sec.  2.1202 by revising paragraphs 
(a)(1) through (3), and (a)(6) to include references to part 53.

Sec.  2.1301 Public Notice of Receipt of a License Transfer Application

    This proposed rule would revise Sec.  2.1301(b) to include a 
corresponding reference to license transfers under part 53 in addition 
to parts 50 and 52.

Sec.  2.1403 Authority and Role of the NRC Staff

    This proposed role would update Sec.  2.1403 to specify that 
``significant hazards considerations'' has the same meaning as defined 
in part 53.

Sec.  2.1500 Purpose and Scope

    This proposed rule would revise Sec.  2.1500 to extend the scope of 
subpart O of part 2 to DC rulemaking hearings under part 53.

Sec.  2.1502 Commission Decision To Hold Legislative Hearing

    This proposed rule would revise Sec.  2.1502, paragraphs (a) and 
(b)(1) to

[[Page 86990]]

incorporate references to part 53 regarding the Commission's decision 
to hold a DC rulemaking.

Sec.  10.1 Purpose

    This proposed rule would revise Sec.  10.1(a)(3) to include a 
reference to part 53.

Sec.  10.2 Scope

    This proposed rule would revise Sec.  10.2(b) to extend the scope 
of subpart A to applicants and holders of licenses, certificates, and 
standard design approvals under part 53 in addition to part 52.

Sec.  11.7 Definitions

    This proposed rule would revise Sec.  11.7 such that terms defined 
in part 53 have the same meaning when used in part 11.

Sec.  19.2 Scope

    This proposed rule would revise Sec.  19.2(a) to include references 
to part 53.

Sec.  19.3 Definitions

    This proposed rule would revise the definitions of ``License'' and 
``Regulated entities'' in Sec.  19.3 to incorporate references to part 
53.

Sec.  19.11 Posting of Notices to Workers

    This proposed rule would amend Sec.  19.11 by revising paragraphs 
(a), (b), and (e)(1) to apply to applicants and holders of licenses, 
permits, standard design approvals, and standard DCs under part 53 in 
addition to part 52.

Sec.  19.14 Presence of Representatives of Licensees and Regulated 
Entities, and Workers During Inspections

    This proposed rule would revise Sec.  19.14(a) to apply to 
applicants and holders of a license, standard design approval, ESP, or 
standard DC under part 53 in addition to part 52.

Sec.  19.20 Employee Protection

    This proposed rule would revise Sec.  19.20 to include a reference 
to protected activities under part 53.

Sec.  20.1002 Scope

    This proposed rule would revise the first sentence of 10 CFR part 
20, ``Standards for Protection Against Radiation,'' Sec.  20.1002 to 
extend the scope of part 20 to apply to persons licensed by the 
Commission to receive, use, transfer, or dispose of byproduct, source, 
or SNM or to operate a production or utilization facility under part 
53.

Sec.  20.1003 Definitions

    This proposed rule would revise Sec.  20.1003 to update the 
definition of ``License'' to include those issued under part 53.

Sec.  20.1101 Radiation Protection Programs

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

Sec.  20.1401 General Provisions and Scope

    This proposed rule would revise Sec.  20.1401, paragraphs (a) and 
(c) to extend the scope of subpart E of part 20 to apply to the 
decommissioning of facilities licensed under part 53 and the release of 
part of a facility or site for unrestricted use in accordance with 
Sec.  53.1080.

Sec.  20.1403 Criteria for License Termination Under Restricted 
Conditions

    This proposed rule would revise Sec.  20.1403(d) to include 
decommissioning plans under part 53.

Sec.  20.1404 Alternate Criteria for License Termination

    This proposed rule would revise Sec.  20.1404(a)(4) to include a 
reference to part 53 regarding alternate criteria for license 
termination.

Sec.  20.1406 Minimization of Contamination

    This proposed rule would revise Sec.  20.1406(a) to include 
references to applicants for licenses other than ESPs or MLs under part 
53. It would also revise Sec.  20.1406(b) to include references to 
standard DCs and standard design approvals under part 53 in addition to 
part 52.

Sec.  20.1501 General

    This proposed rule would revise Sec.  20.1501(b) regarding the 
requirement for retention of records from surveys describing the 
location and amount of subsurface residual radioactivity at a site to 
include a reference to the retention requirements under part 53.

Sec.  20.1905 Exemptions to Labeling Requirements

    This proposed rule would revise Sec.  20.1905(g) to apply to 
facilities licensed under part 53 in addition to parts 50 and 52 
regarding exemptions to labeling requirements.

Sec.  20.2004 Treatment or Disposal by Incineration

    This proposed rule would revise Sec.  20.2004(b)(1) to include 
references to part 53 regarding the treatment or disposal of waste oil 
by incineration.

Sec.  20.2201 Reports of Theft or Loss of Licensed Material

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

Sec.  20.2202 Notification of Incidents

    This proposed rule would revise Sec.  20.2202(d)(1) to add 
references to part 53 regarding reports to the NRC Operations Center.

Sec.  20.2203 Reports of Exposures, Radiation Levels, and 
Concentrations of Radioactive Material Exceeding the Constraints or 
Limits

    This proposed rule would revise Sec.  20.2203(c) to refer to 
procedures under part 53 for reporting occurrences of exposures, 
radiation levels, and concentrations of radioactive material exceeding 
the constraints or limits.

Sec.  20.2206 Reports of Individual Monitoring

    This proposed rule would revise Sec.  20.2206(a)(1) to include a 
reference to part 53.

Sec.  21.2 Scope

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

Sec.  21.3 Definitions

    This proposed rule, in Sec.  21.3 would revise the definitions of 
``Basic component,'' ``Commercial grade item,'' ``Critical 
characteristics,'' ``Dedicating entity,'' ``Dedication,'' ``Defect,'' 
and ``Substantial safety hazard'' with references to part 53.

Sec.  21.21 Notification of Failure To Comply or Existence of a Defect 
and Its Evaluation

    This proposed rule would revise Sec.  21.21, by incorporating 
references to part 53, to update the requirements for notifying the 
Commission of a failure to comply or defect in paragraphs (a)(3) and 
(d)(1).

Sec.  21.51 Maintenance and Inspection of Records

    This proposed rule would revise Sec.  21.51(a)(4) and (5) to apply 
to applicants for standard DC and applicants or holders of a standard 
design approval under part 53, in addition to part 52, regarding the 
retention of records.

[[Page 86991]]

Sec.  21.61 Failure To Notify

    This proposed rule would revise Sec.  21.61(b) to include 
references to part 53 licensees and applicants regarding failure to 
provide the notice required in Sec.  21.21.

Sec.  25.5 Definitions

    This proposed rule would update the definition of ``License'' to 
include those issued under part 53.

Sec.  25.17 Approval for Processing Applicants for Access Authorization

    This proposed rule would revise Sec.  25.17(a) to add a reference 
to part 53 regarding AAs for individuals who need access to classified 
information in connection with activities under part 53.

Sec.  25.35 Classified Visits

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

Sec.  26.3 Scope

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

Sec.  26.4 FFD Program Applicability to Categories of Individuals

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

Sec.  26.5 Definitions

    This proposed rule would amend Sec.  26.5 by adding definitions for 
``Biological marker,'' ``Change,'' ``Illicit substance,'' ``Reduction 
in FFD program effectiveness,'' and ``Special Nuclear Material.'' It 
would also revise definitions of ``Constructing or construction 
activities,'' ``Contractor/vendor (C/V),'' ``Other entity,'' 
``Questionable validity,'' ``Reviewing official,'' ``Safety-related 
structures, systems, and components (SSCs),'' ``Security-related 
SSCs,'' and ``Unit outage'' within this section.

Sec.  26.8 Information Collection Requirements: OMB Approval

    This proposed rule would revise Sec.  26.8(b) with the new 
information collection requirements contained in proposed Sec. Sec.  
26.202, 26.603, 26.604, 26.605, 26.606, 26.607, 26.608, 26.609, 26.611, 
26.613, 26.617, and 26.619.

Sec.  26.21 Fitness-for-Duty Program

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

Sec.  26.51 Applicability

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

Sec.  26.53 General Provisions

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

Sec.  26.63 Suitable Inquiry

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

Sec.  26.73 Applicability

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

Sec.  26.81 Purpose and Applicability

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

Sec.  26.201 Applicability

    This proposed rule would revise Sec.  26.201 to include references 
to the proposed provisions in Sec. Sec.  26.3(f) and 26.202, as well as 
revise the applicability of requirements in subpart I of part 26.

Sec.  26.202 General Provisions for Facilities Licensed Under Part 53

    This proposed rule would add new Sec.  26.202, which would require 
applicable licensees under part 53 to incorporate a policy for fatigue 
management into their FFD program in accordance with the provisions of 
this section.

Sec.  26.205 Work Hours

    This proposed rule would revise paragraphs (d)(7)(iii) and (d)(8) 
of Sec.  26.205 to incorporate references to Sec. Sec.  26.606 and 
26.202(a) and (b).

Sec.  26.207 Waivers and Exceptions

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

Sec.  26.211 Fatigue Assessments

    This proposed rule would revise Sec.  26.211, paragraphs (a)(1), 
(a)(3), and (b) to incorporate references to Sec. Sec.  26.202(c), 
26.607(b), 26.608, and 26.619 and to include provisions for 
implementing certain face-to-face assessments using electronic 
communications.

Subpart M--Fitness for Duty Programs for Facilities Licensed Under Part 
53

    This proposed rule would add new Subpart M of part 26 containing 
Sec. Sec.  26.601, 26.603, 26.604 through 26.611, 26.613, 26.615, 
26.617, and 26.619, which adds an optional technology-inclusive, risk-
informed, and performance-based approach for the application of drug 
and alcohol testing and fatigue management requirements for facilities 
licensed under part 53.

Sec.  26.601 Applicability

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

Sec.  26.603 General Provisions

    This proposed rule would add Sec.  26.603, which would establish 
the general requirements for implementing an FFD program under subpart 
M of part 26.

Sec.  26.604 FFD Program Requirements for Facilities That Satisfy the 
Sec.  26.603(c) Criterion

    This proposed rule would add Sec.  26.604, which would establish 
the FFD program elements for a licensee or other entity whose 
facilities and operations demonstrate compliance with the criterion in 
Sec.  26.603(c).

Sec.  26.605 FFD Program Requirements for Facilities That Do Not 
Implement Sec.  26.604

    This proposed rule would add Sec.  26.605, which would establish 
the FFD program elements for a licensee or other entity that does not 
demonstrate compliance with the criterion in Sec.  26.603(c), or 
otherwise chooses to maintain an FFD program under this section.

[[Page 86992]]

Sec.  26.606 Written Policies and Procedures

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

Sec.  26.607 Drug and Alcohol Testing

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

Sec.  26.608 FFD Program Training

    This proposed rule would add Sec.  26.608, which would require 
individuals who are subject to the FFD program under subpart M of part 
26 to receive periodic training on FFD policies and procedures, 
including their duties and responsibilities under the BOP.

Sec.  26.609 Behavioral Observation

    This proposed rule would add Sec.  26.609, which would establish 
the requirements for a BOP under subpart M of part 26.

Sec.  26.610 Sanctions

    This proposed rule would add Sec.  26.610, which would require 
licensees and other entities implementing an FFD program under subpart 
M of part 26 to establish sanctions for FFD policy violations.

Sec.  26.611 Protection of Information

    This proposed rule would add Sec.  26.611, which would require 
licensees and other entities implementing an FFD program under subpart 
M of part 26 to establish a system to protect personal information 
against unauthorized disclosure.

Sec.  26.613 Appeals Process

    This proposed rule would add Sec.  26.613, which would require 
licensees and other entities that implement an FFD program under 
subpart M of part 26 to establish procedures for an individual to 
appeal a policy violation determination.

Sec.  26.615 Audits

    This proposed rule would add Sec.  26.615, which would establish 
provisions for licensees and other entities that implement an FFD 
program under subpart M of part 26 to conduct audits to monitor the 
effectiveness of FFD program elements.

Sec.  26.617 Recordkeeping and Reporting

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

Sec.  26.619 Suitability and Fitness Determinations

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

Sec.  26.709 Applicability

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

Sec.  26.711 General Provisions

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

Sec.  26.825 Criminal Penalties

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

Sec.  30.4 Definitions

    This proposed rule would revise the definition for ``Utilization 
facility'' in Sec.  30.4 to include utilization facilities defined in 
the regulations under part 53 in addition to part 50.

Sec.  30.50 Reporting Requirements

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

Sec.  40.60 Reporting Requirements

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

Sec.  50.47 Emergency Plans

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

Sec.  50.54 Conditions of Licenses

    This proposed rule would revise Sec.  50.54(q)(2), (q)(4), and 
(gg)(1) with appropriate references to part 53.

Sec.  50.160 Emergency Preparedness for Small Modular Reactors, Non-
Light-Water Reactors, and Non-Power Production or Utilization 
Facilities

    This proposed rule would revise Sec.  50.160(b)(3) and (c)(2) with 
the appropriate references to part 53.

Appendix B to 10 CFR Part 50--Quality Assurance Criteria for Nuclear 
Power Plants and Fuel Reprocessing Plants

    This proposed rule would revise appendix B to part 50 by revising 
the introduction and specific criteria to incorporate the appropriate 
references and terminology for part 53.

Sec.  51.20 Criteria for and Identification of Licensing and Regulatory 
Actions Requiring Environmental Impact Statements

    This proposed rule would revise Sec.  51.20(b)(1) and (2) to 
require an EIS prior to the issuance of a CP, LWA, or ESP under part 
53, or the issuance to renewal of a full power or design capacity 
license to operate a nuclear power reactor, testing facility, or fuel 
reprocessing plant under part 53.

Sec.  51.22 Criterion for Categorical Exclusion; Identification of 
Licensing and Regulatory Actions Eligible for Categorical Exclusion or 
Otherwise Not Requiring Environmental Review

    This proposed rule would revise Sec.  51.22 to include 
corresponding references to part 53 in paragraphs (c)(3), (c)(9), 
(c)(12), (c)(17), (c)(22) and (23).

Sec.  51.26 Requirement To Publish Notice of Intent and Conduct Scoping 
Process

    This proposed rule would revise Sec.  51.26(d) to add a reference 
to part 53.

Sec.  51.30 Environmental Assessment

    This proposed rule would revise the introductory text to paragraph 
(a) and revise paragraphs (d) and (e) of Sec.  51.30 to incorporate the 
appropriate references to part 53 regarding EAs.

Sec.  51.31 Determinations Based on Environmental Assessment

    This proposed rule would revise Sec.  51.31(a) to include a 
reference to part 53.

[[Page 86993]]

Sec.  51.32 Finding of No Significant Impact

    This proposed rule would revise Sec.  51.32(b)(1) and (3), finding 
there is no significant environmental impact associated with the 
issuance of standard DCs and MLs under part 53.

Sec.  51.49 Environmental Report-Limited Work Authorization

    This proposed rule would revise the introductory text of Sec.  
51.49(c) to require applicants for an ESP under part 53 requesting a 
LWA to include the environmental report required by Sec.  51.50(b).

Sec.  51.50 Environmental Report--Construction Permit, Early Site 
Permit, or Combined License Stage

    This proposed rule would revise Sec.  51.50, paragraphs (a), 
(b)(4), and the introductory text for paragraph (c) to incorporate the 
appropriate references to part 53.

Sec.  51.53 Postconstruction Environmental Reports

    This proposed rule would revise Sec.  51.53(d) to include the 
appropriate references to part 53 regarding a license termination plan 
or decommissioning plan and related requirements for postconstruction 
environmental reports.

Sec.  51.54 Environmental Report--Manufacturing License

    This proposed rule would update Sec.  51.54(a) to require 
applicants for MLs under part 53 to submit an environmental report with 
the application.

Sec.  51.55 Environmental Report--Standard Design Certification

    This proposed rule would update Sec.  51.55(a) to require 
applicants for a standard DC under part 53 to submit an environmental 
report with the application.

Sec.  51.58 Environmental Report--Number of Copies; Distribution

    This proposed rule would revise Sec.  51.58(b) to incorporate the 
appropriate references to part 53.

Sec.  51.77 Distribution of Draft Environmental Impact Statement

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

Sec.  51.92 Supplement to the Final Environmental Impact Statement

    This proposed rule would revise Sec.  51.92(b) to apply to COL 
applications referencing an ESP under part 53.

Sec.  51.95 Postconstruction Environmental Impact Statements

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

Sec.  51.101 Limitations on Actions

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

Sec.  51.103 Record of Decision--General

    This proposed rule would update Sec.  51.103(a)(6) to apply to the 
issuance of a LWA in connection with a CP or COL under part 53.

Sec.  51.105 Public Hearings in Proceedings for Issuance of 
Construction Permits or Early Site Permits; Limited Work Authorizations

    This proposed rule would revise Sec.  51.105(c)(1) to include the 
appropriate reference to LWAs under part 53 for CPs or ESPs.

Sec.  51.107 Public Hearings in Proceedings for Issuance of Combined 
Licenses; Limited Work Authorizations

    This proposed rule would amend Sec.  51.107 by revising the 
introductory text for paragraphs (a) and (b) and updating paragraph 
(d)(1) to include the appropriate corresponding references to part 53.

Sec.  51.108 Public Hearings on Commission Findings That Inspections, 
Tests, Analyses, and Acceptance Criteria of Combined Licenses Are Met

    This proposed rule would revise Sec.  51.108 to incorporate the 
appropriate references to part 53.

10 CFR part 53--Risk-Informed, Technology-Inclusive Regulatory 
Framework for Commercial Nuclear Plants

    This proposed rule would add a new part to 10 CFR Chapter I, 
designated as Part 53 including Sec. Sec.  53.000 through 53.9010.

Sec.  53.000 Purpose

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

Subpart A--General Provisions

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

Sec.  53.015 Scope

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

Sec.  53.020 Definitions

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

Sec.  53.040 Written Communications

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

Sec.  53.050 Deliberate Misconduct

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

Sec.  53.060 Employee Protection

    This proposed rule would add Sec.  53.060, which would prohibit 
applicants and licensees from discriminating against employees for 
engaging in the protected activities listed in this section and provide 
remedial procedures for employees who believe they are the subjects of 
discrimination.

Sec.  53.070 Completeness and Accuracy of Information

    This proposed rule would add Sec.  53.070, which would require 
licensees and applicants under part 53 to provide complete and accurate 
information in accordance with all applicable laws, Commission 
regulations, and the terms and conditions of their license. This 
proposed rule would also require licensees to notify the Commission 
within two days of identifying information with material implications

[[Page 86994]]

for public health and safety or common defense and security.

Sec.  53.080 Specific Exemptions

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

Sec.  53.090 Standards for Review

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

Sec.  53.100 Jurisdictional Limits

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

Sec.  53.110 Attacks and Destructive Acts

    This proposed rule would add Sec.  53.110, which would exempt 
licensees or applicants under part 53 from providing design features to 
protect against attacks or destructive acts directed at the facility by 
United States adversaries.

Sec.  53.115 Rights Related to Special Nuclear Material

    This proposed rule would add Sec.  53.115, which would establish 
provisions regarding the rights to SNM under a part 53 license.

Sec.  53.117 License Suspension and Rights of Recapture

    This proposed rule would add Sec.  53.117, which would provide that 
the Commission may suspend licenses and recapture material or control 
of a facility in a state of war or national emergency declared by 
Congress.

Sec.  53.120 Information Collection Requirements: OMB Approval

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

Subpart B--Technology-Inclusive Safety Requirements

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

Sec.  53.210 Safety Criteria for Design-Basis Accidents

    This proposed rule would add Sec.  53.210 to set dose values to 
ensure that plants are designed to limit the public's radiation 
exposure in the event of a DBA.

Sec.  53.220 Safety Criteria for Licensing-Basis Events Other Than 
Design-Basis Accidents

    This proposed rule would add Sec.  53.220 to require plants to 
implement a combination of design features and programmatic controls to 
control risks to the public in the event of a LBE other than a DBA.

Sec.  53.230 Safety Functions

    This proposed rule would add Sec.  53.230, which specifies that 
limiting the release of radioactive materials from the facility is the 
primary safety function of a commercial nuclear plant, and that 
additional safety functions must be defined to support the retention of 
radioactive materials during LBEs.

Sec.  53.240 Licensing-Basis Events

    This proposed rule would add Sec.  53.240 to require commercial 
nuclear plants to conduct an analysis of LBEs to confirm that design 
features and programmatic controls satisfy the safety criteria under 
Sec. Sec.  53.210 and 53.220, or alternatively, under Sec.  53.470.

Sec.  53.250 Defense in Depth

    This proposed rule would add Sec.  53.250 to establish a 
performance-based, defense-in-depth approach to address uncertainties 
about the effectiveness and reliability of plant SSCs, personnel, and 
programmatic controls.

Sec.  53.260 Normal Operations

    This proposed rule would add Sec.  53.260, requiring holders of 
licenses to operate commercial nuclear plants to control public doses 
and dose rates in unrestricted areas to meet the requirements in part 
20, during normal plant operation.

Sec.  53.270 Protection of Plant Workers

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

Subpart C--Design and Analysis Requirements

    This proposed rule would add subpart C, which requires the 
implementation of certain design features and the performance of risk 
assessments and analyses to demonstrate compliance with the safety 
criteria and safety functions in subpart B.

Sec.  53.400 Design Features for Licensing-Basis Events

    This proposed rule would add Sec.  53.400, which would require 
design features that satisfy the safety criteria defined in Sec.  
53.210 and Sec.  53.220 or Sec.  53.470 and fulfill the safety 
functions identified in Sec.  53.230 during LBEs.

Sec.  53.410 Functional Design Criteria for Design-Basis Accidents

    This proposed rule would add Sec.  53.410, which would stipulate 
that functional design criteria must be defined for each design feature 
required by Sec.  53.400 to demonstrate compliance with the safety 
criteria defined in Sec.  53.210 for DBAs.

Sec.  53.415 Protection Against External Hazards

    This proposed rule would add Sec.  53.415, which would require SR 
SSCs to be designed to withstand the effects of natural phenomena and 
constructed hazards while performing the intended safety functions.

Sec.  53.420 Functional Design Criteria for Licensing-Basis Events 
Other Than Design-Basis Accidents

    This proposed rule would add Sec.  53.420, which would require 
functional design criteria to be defined for each design feature 
required by Sec.  53.400 to demonstrate compliance with the safety 
criteria defined in Sec.  53.220 for LBEs other than DBAs.

Sec.  53.425 Design Features and Functional Design Criteria for Normal 
Operations

    This proposed rule would add Sec.  53.425, which would require 
commercial nuclear plants to implement design features and define 
functional design criteria sufficient to demonstrate compliance with 
Sec.  53.850 and show through functional design criteria that design 
features and corresponding programmatic controls control wastes, as 
required under part 20.

Sec.  53.430 Design Features and Functional Design Criteria for 
Protection of Plant Workers

    This proposed rule would add Sec.  53.430, which would require 
commercial nuclear plants to implement design features and define 
functional design criteria sufficient to demonstrate compliance with 
Sec.  53.270.

Sec.  53.440 Design Requirements

    This proposed rule would add Sec.  53.440, which would establish 
various

[[Page 86995]]

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

Sec.  53.450 Analysis Requirements

    This proposed rule would add Sec.  53.450, which would require 
commercial nuclear plants to perform PRAs in combination with other 
analytical methods to identify and assess risks and determine 
compliance with the safety criteria in subpart B. In addition, Sec.  
53.450 would require analysis of DBAs and other analyses to assess the 
adequacy of protections against fire, aircraft impact, and the release 
of effluents.

Sec.  53.460 Safety Categorization and Special Treatments

    This proposed rule would add Sec.  53.460 to address the safety 
classification of SSCs and determine appropriate special treatments.

Sec.  53.470 Maintaining Analytical Safety Margins Used To Justify 
Operational Flexibilities

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

Sec.  53.480 Earthquake Engineering

    This proposed rule would add Sec.  53.480 to provide overall 
seismic design considerations based on the safety criteria in subpart B 
and siting requirements in subpart D to ensure that SSCs are able to 
withstand the effects of earthquakes without loss of capability to 
fulfill safety functions.

Subpart D--Siting Requirements

    This proposed rule would add subpart D, which would address 
requirements associated with the siting of commercial nuclear 
facilities under part 53, including considerations of external hazards 
and potential adverse impacts on the surrounding population.

Sec.  53.500 General Siting and Siting Assessment

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

Sec.  53.510 External Hazards

    This proposed rule would add Sec.  53.510, which would require 
site-specific assessments, including an evaluation of geological and 
seismic siting factors, to identify and characterize the external 
hazard level for a range of natural and constructed hazards.

Sec.  53.520 Site Characteristics

    This proposed rule would add Sec.  53.520, which would require the 
design and analyses conducted under subpart C to consider how site 
characteristics may contribute to LBEs.

Sec.  53.530 Population-Related Considerations

    This proposed rule would add Sec.  53.530, which would establish 
requirements related to the facility's exclusion area, low-population 
zone, and population center distance.

Sec.  53.540 Siting Interfaces

    This proposed rule would add Sec.  53.540, which would require that 
external hazards and site characteristics must be accounted for in the 
design features, programmatic controls, and supporting analyses used to 
demonstrate compliance with the safety criteria in Sec. Sec.  53.210 
and 53.220.

Subpart E--Construction and Manufacturing Requirements

    This proposed rule would add subpart E, which would establish 
requirements for the construction and manufacture of commercial nuclear 
plants.

Sec.  53.600 Construction and Manufacturing--Scope and Purpose

    This proposed rule would add Sec.  53.600, which would indicate 
that this subpart applies to construction and manufacturing activities 
authorized by a CP, COL, ML, or LWA issued under this part.

Sec.  53.605 Reporting of Defects and Noncompliance

    This proposed rule would add Sec.  53.605, which would describe the 
procedures, notification requirements, and records retention 
requirements that each CP, ML, and COL is subject to with respect to 
reporting of defects and noncompliance.

Sec.  53.610 Construction

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

Sec.  53.620 Manufacturing

    This proposed rule would add Sec.  53.620, which would ensure that 
the holders of an ML under part 53 develop plans, programs, and 
organizational units to manage and control manufacturing activities, 
and would establish requirements for the loading of fuel into a 
manufactured reactor for subsequent transport to a commercial nuclear 
plant and operation pursuant to a COL.

Subpart F--Requirements for Operation

    This proposed rule would add subpart F, which would establish 
regulatory requirements to ensure that the safety criteria in subpart B 
are satisfied whenever a commercial nuclear plant licensed under part 
53 is operational. This includes periods of normal operation and 
unplanned events.

Sec.  53.700 Operational Objectives

    This proposed rule would add Sec.  53.700, which would establish 
general operational objectives to ensure that licensees under part 53 
have implemented and maintained the SSCs necessary to demonstrate 
compliance with the safety functions identified in subpart B for 
addressing normal operations and responding to LBEs.

Sec.  53.710 Maintaining Capabilities and Availability of Structures, 
Systems, and Components

    This proposed rule would add Sec.  53.710, which would require 
licensees under part 53 to demonstrate compliance with the safety 
criteria in subpart B by establishing TS for all SR SSCs and developing 
documents and procedures for all NSRSS SSCs.

Sec.  53.715 Maintenance, Repair, and Inspection Programs

    This proposed rule would add Sec.  53.715, which would require 
licensees to develop, implement, and maintain programs to assess and 
manage any risks posed by maintenance activities and to evaluate the 
efficacy of performance, condition monitoring, and maintenance 
activities.

Sec.  53.720 Response to Seismic Events

    This proposed rule would add Sec.  53.720, which would establish 
requirements for licensees to respond to a seismic event during the 
operating phase of the life cycle of a commercial nuclear plant.

Sec.  53.725 General Staffing, Training, Personnel Qualifications, and 
Human Factors Requirements

    This proposed rule would add Sec.  53.725, which would provide an

[[Page 86996]]

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

Sec.  53.726 Communications

    This proposed rule would add Sec.  53.726, which would contain 
communications requirements applicable to sections Sec. Sec.  53.725 
through 53.830. It also contains requirements to notify the Commission 
within 30 days should a specifically licensed operator or senior 
operator be reassigned, terminated, or suffer permanent disability or 
illness.

Sec.  53.728 Completeness and Accuracy of Information

    This proposed rule would add Sec.  53.728, which would require 
submitted information to be complete and accurate in all material 
respects.

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

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

Sec.  53.735 General Exemptions

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

Sec.  53.740 Facility Licensee Requirements--General

    This proposed rule would add Sec.  53.740, which would establish 
staffing requirements for interaction-dependent-mitigation facilities 
and self-reliant mitigation facilities.

Sec.  53.745 Operator License Requirements

    This proposed rule would add Sec.  53.745, which would require 
individuals to be licensed to perform certain functions.

Sec.  53.760 Operator Licensing

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

Sec.  53.765 Medical Requirements

    This proposed rule would add Sec.  53.765, which would establish 
medical requirements for specifically licensed operators and senior 
operators.

Sec.  53.770 Incapacitation Because of Disability or Illness

    This proposed rule would add Sec.  53.770, which would establish 
requirements to address permanent medical conditions for specifically 
licensed operators and senior operators.

Sec.  53.775 Applications for Operators and Senior Operators

    This proposed rule would add Sec.  53.775, which would establish 
the application process and requirements for individuals applying for 
specific operator and senior operator licenses.

Sec.  53.780 Training, Examination, and Proficiency Program

    This proposed rule would add Sec.  53.780, which would contain the 
requirements associated with specifically licensed operator and senior 
operator initial training, initial examinations, requalification 
training, requalification examinations, examination integrity, 
simulation facilities, waivers, and proficiency.

Sec.  53.785 Conditions of Operator and Senior Operator Licenses

    This proposed rule would add Sec.  53.785, which would establish 
conditions for specific operator and senior operator licenses.

Sec.  53.790 Issuance, Modification, and Revocation of Operator and 
Senior Operator Licenses

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

Sec.  53.795 Expiration and Renewal of Operator and Senior Operator 
Licenses

    This proposed rule would add Sec.  53.795, which would contain 
requirements associated with the expiration and renewal of specific 
operator and senior operator licenses.

Sec.  53.800 Facility Licensees for Self-Reliant-Mitigation Facilities

    This proposed rule would add Sec.  53.800, which would establish 
the technical criteria by which commercial nuclear plants under part 53 
are determined to be of the self-reliant mitigation class of facilities 
that would be staffed by GLROs in lieu of specifically licensed 
operators and senior operators.

Sec.  53.805 Facility Licensee Requirements Related to Generally 
Licensed Reactor Operators

    This proposed rule would add Sec.  53.805, which would establish 
requirements that apply to the facility licensee at those facilities 
staffed by GLROs.

Sec.  53.810 Generally Licensed Reactor Operators

    This proposed rule would add Sec.  53.810, which would issue and 
describe the general license for GLROs that manipulate the controls of 
a self-reliant mitigation facility.

Sec.  53.815 Generally Licensed Reactor Operator Training, Examination, 
and Proficiency Programs

    This proposed rule would add Sec.  53.815, which would contain the 
requirements for GLRO initial training, initial examinations, 
continuing training, requalification examinations, examination 
integrity, simulation facilities, examination waivers, and proficiency.

Sec.  53.820 Cessation of Individual Applicability

    This proposed rule would add Sec.  53.820, which would address the 
requirements by which the general license for GLROs would cease to be 
applicable on an individual basis.

Sec.  53.830 Training and Qualification of Commercial Nuclear Plant 
Personnel

    This proposed rule would add Sec.  53.830, which would address 
training and qualification requirements for supervisors, technicians, 
and other appropriate operating personnel at commercial nuclear plants.

Sec.  53.845 Programs

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

[[Page 86997]]

B are maintained during normal operations and LBEs.

Sec.  53.850 Radiation Protection

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

Sec.  53.855 Emergency Preparedness

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

Sec.  53.860 Security Programs

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

Sec.  53.865 Quality Assurance

    This proposed rule would add Sec.  53.865, which would require 
licensees under part 53 to establish a quality assurance program that 
includes a written manual to ensure activities are conducted in 
accordance with codes and standards found acceptable by the NRC.

Sec.  53.870 Integrity Assessment Programs

    This proposed rule would add Sec.  53.870, which would require 
licensees under part 53 to establish an integrity assessment program to 
ensure that the plant continues to fulfill safety criteria and 
functional design criteria as it ages.

Sec.  53.875 Fire Protection

    This proposed rule would add Sec.  53.875, which would require 
licensees under part 53 to establish a fire protection plan and 
describe the necessary elements that the plan must incorporate.

Sec.  53.880 Inservice Inspection and Inservice Testing

    This proposed rule would add Sec.  53.880, which would require 
licensees under part 53 to develop and implement a program for ISI and 
IST in accordance with the requirements of this section.

Sec.  53.910 Procedures and Guidelines

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

Subpart G--Decommissioning Requirements

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

Sec.  53.1000 Scope and Purpose

    This proposed rule would add Sec.  53.1000, which would establish 
the scope of the decommissioning requirements for applicants and 
licensees under part 53 and describe the contents of subpart G of part 
53.

Sec.  53.1010 Financial Assurance for Decommissioning

    This proposed rule would add Sec.  53.1010, which would establish 
the requirement that applicants for an OL or COL under part 53 provide 
reasonable assurance that funds will be available for the 
decommissioning process. This section would describe the requirements 
associated with the required plan and an associated decommissioning 
report that ensures and documents that adequate funding for 
decommissioning will be available.

Sec.  53.1020 Cost Estimates for Decommissioning

    This proposed rule would add Sec.  53.1020, which would require 
site-specific cost estimates for decommissioning and establish the 
aspects that must be included in the estimate.

Sec.  53.1030 Annual Adjustments to Cost Estimates for Decommissioning

    This proposed rule would add Sec.  53.1030, which would require 
that holders of an OL or COL under part 53 annually adjust their cost 
estimate for decommissioning to account for escalation in labor, 
energy, and waste burial costs. This section would allow licensees to 
elect either a site-specific adjustment factor or a generic adjustment 
factor.

Sec.  53.1040 Methods for Providing Financial Assurance for 
Decommissioning

    This proposed rule would add Sec.  53.1040, which would establish 
suitable methods that holders of an OL or COL under part 53 may use to 
provide financial assurance for decommissioning to the NRC.

Sec.  53.1045 Limitations on the Use of Decommissioning Trust Funds

    This proposed rule would add Sec.  53.1045, which would establish 
requirements for decommissioning trust funds under part 53, including 
criteria for using decommissioning trust funds and required terms.

Sec.  53.1050 NRC Oversight

    This proposed rule would add Sec.  53.1050, which would outline the 
steps the NRC may take to ensure adequate accumulation of 
decommissioning funds.

Sec.  53.1060 Reporting and Recordkeeping Requirements

    This proposed rule would add Sec.  53.1060, which would contain 
reporting and recordkeeping requirements related to decommissioning for 
each holder of an OL or COL under part 53. This section would outline 
requirements for documents such as: certification of decommissioning 
funding, decommissioning cost estimates and copies of financial 
instruments, licensee records of information important to safe and 
effective decommissioning, post-shutdown decommissioning activities 
report, financial assurance reports, and reports on the status of 
funding for managing irradiated fuel.

Sec.  53.1070 Termination of License

    This proposed rule would add Sec.  53.1070, which would establish 
procedures for decommissioning and license termination applicable to 
licensees under part 53 that have determined to permanently cease 
operations.

Sec.  53.1075 Program Requirements During Decommissioning

    This proposed rule would add Sec.  53.1075, which would require 
licensees under part 53 to establish and maintain a decommissioning 
fire protection program to prevent, detect, and control fires, and 
ensure that the risk of fire induced radiological hazards are minimized 
through the various stages of facility decommissioning.

Sec.  53.1080 Release of Part of a Commercial Nuclear Plant or Site for 
Unrestricted Use

    This proposed rule would add Sec.  53.1080, which would establish 
licensee procedures for requesting and NRC procedures for approving 
partial release of a commercial nuclear plant or site for unrestricted 
use prior to receiving approval of a license termination plan from the 
Commission under part 53.

Subpart H--Licenses, Certifications, and Approvals

    This proposed rule would add subpart H, which would govern the 
process of applying for, amending, renewing, or

[[Page 86998]]

terminating a LWA, ESP, standard design approval, standard DC, ML, CP, 
OL, or COL under part 53.

Sec.  53.1100 Filling of Application for Licenses, Certifications, or 
Approvals; Oath or Affirmation

    This proposed rule would add Sec.  53.1100, which would establish 
requirements for applicants seeking a standard design approval, 
standard DC, license, or permit under part 53 to submit an application.

Sec.  53.1101 Requirement for License

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

Sec.  53.1103 Combining Applications and Licenses

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

Sec.  53.1106 Elimination of Repetition

    This proposed rule would add Sec.  53.1106, which would allow 
applicants under part 53 to reference information contained in previous 
documents filed with the Commission so long as those references are 
clear and specific.

Sec.  53.1109 Contents of Applications; General Information

    This proposed rule would add Sec.  53.1109, which would establish 
the general content to be included in applications made under part 53, 
including but not limited to the identifying information of the 
applicant and the radiological emergency response plans of government 
entities within the plume exposure pathway EPZ.

Sec.  53.1112 Environmental Conditions

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

Sec.  53.1115 Agreement Limiting Access to Classified Information

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

Sec.  53.1118 Ineligibility of Certain Applicants

    This proposed rule would add Sec.  53.1118, which would prevent 
citizens, nationals, or agents of a foreign country or corporations 
owned, controlled, or dominated by a foreign entity from applying for 
or obtaining a license under part 53.

Sec.  53.1120 Exceptions and Exemptions From Licensing Requirements

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

Sec.  53.1121 Public Inspection of Applications

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

Sec.  53.1124 Relationship Between Sections

    This proposed rule would add Sec.  53.1124, which would outline the 
relationship between LWAs, ESPs, standard design approvals, standard 
DCs, MLs, CPs, OLs, and COLs under part 53.

Sec.  53.1130 Limited Work Authorizations

    This proposed rule would add Sec.  53.1130, which would establish 
requirements for requesting an LWA and grounds for the Commission to 
issue an LWA. It would also contain details about the effect of an LWA 
and the implementation of a redress plan.

Sec.  53.1140 Early Site Permits

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

Sec.  53.1143 Filing of Applications

    This proposed rule would add Sec.  53.1143, which would enable an 
applicant under part 53 to apply for an ESP, regardless of whether they 
have filed an application for a CP or COL for that site.

Sec.  53.1144 Contents of Applications for Early Site Permits; General 
Information

    This proposed rule would add Sec.  53.1144, which would require 
applications for ESPs to include the information required by Sec.  
53.1109(a) through (d) and (j).

Sec.  53.1146 Contents of Applications for Early Site Permits; 
Technical Information

    This proposed rule would add Sec.  53.1146, which would require 
applicants for ESPs to submit technical information, including but not 
limited to a Site Safety Analysis Report and emergency plans.

Sec.  53.1149 Review of Applications

    This proposed rule would add Sec.  53.1149, which would establish 
standards for review of applications for ESPs under part 53, including 
requirements for the Commission to prepare an EIS and assess the 
adequacy of protective actions in the event of a radiological 
emergency. It would also require the administrative review of 
applications and hearings to follow the procedural requirements of part 
2.

Sec.  53.1155 Referral to the Advisory Committee on Reactor Safeguards

    This proposed rule would add Sec.  53.1115, which would require the 
ACRS to review SR content in the application for an ESP under part 53.

Sec.  53.1158 Issuance of Early Site Permit

    This proposed rule would add Sec.  53.1158, which would establish 
the conditions under which the Commission may issue an ESP under part 
53, as well as the information, terms, and conditions to be included in 
the permit.

Sec.  53.1161 Extent of Activities Permitted

    This proposed rule would add Sec.  53.1161, which would require 
that a valid ESP only be used for the purpose of site redress, unless 
the site is referenced in an application for a CP or COL under part 53.

Sec.  53.1164 Duration of Permit

    This proposed rule would add Sec.  53.1164, which would govern the 
conditions under which an ESP remains valid following the date of 
issuance.

Sec.  53.1167 Limited Work Authorization After Issuance of Early Site 
Permit

    This proposed rule would add Sec.  53.1167, which would permit the

[[Page 86999]]

holder of an ESP to request a LWA under Sec.  53.1130.

Sec.  53.1170 Transfer of Early Site Permit

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

Sec.  53.1173 Application for Renewal

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

Sec.  53.1176 Criteria for Renewal

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

Sec.  53.1179 Duration of Renewal

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

Sec.  53.1182 Use of Site for Other Purposes

    This proposed rule would add Sec.  53.1182, which would govern 
acceptable uses of the site for purposes other than those described in 
the permit.

Sec.  53.1188 Finality of Early Site Permit Determinations

    This proposed rule would add Sec.  53.1188, which would address the 
finality of ESP determinations under part 53.

Sec.  53.1200 Standard Design Approvals

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

Sec.  53.1203 Filing of Applications

    This proposed rule would add Sec.  53.1203, which would enable 
applicants to submit a final design for the entire facility, or major 
portions, to the NRC staff for review.

Sec.  53.1206 Contents of Applications for Standard Design Approvals; 
General Information

    This proposed rule would add Sec.  53.1206, which would require 
applications for a standard design approval under part 53 to contain 
the information required by Sec.  53.1109(a) through (c) and (j).

Sec.  53.1209 Contents of Applications for Standard Design Approvals; 
Technical Information

    This proposed rule would add Sec.  53.1209, which would require the 
inclusion of certain technical information, including a FSAR, site 
parameters, and design information, when an applicant seeks review of 
major portions of a standard design.

Sec.  53.1210 Contents of Applications for Standard Design Approvals; 
Other Application Content

    This proposed rule would add Sec.  53.1210, which would require 
applications for standard design approvals under part 53 to include a 
description of the availability controls used to satisfy the safety 
criteria of Sec.  53.220, the program to protect Safeguards Information 
against unauthorized disclosure, evidence that safety questions 
associated with SSCs have been resolved, and a description of how 
design features fulfill design criteria.

Sec.  53.1212 Standards for Review of Applications

    This proposed rule would add Sec.  53.1212, which would require 
applications for standard design approval to be reviewed under the 
standards in parts 20, 53, and 73.

Sec.  53.1215 Referral to the Advisory Committee on Reactor Safeguards

    This proposed rule would add Sec.  53.1215, which would require the 
ACRS to report on any portions of the application for a standard design 
approval under part 53 concerning safety.

Sec.  53.1218 Staff Approval of Design

    This proposed rule would add Sec.  53.1218, which would require the 
NRC staff to make a determination on the acceptability of the design, 
publish its decision in the Federal Register, and issue a report 
analyzing the design that is available at http://nrc.gov. Additionally, 
the rule would establish the conditions under which a design approval 
under part 53 remains valid.

Sec.  53.1221 Finality of Standard Design Approvals; Information 
Requests

    This proposed rule would add Sec.  53.1221, which would require NRC 
staff and the ACRS to rely upon an approved design in their review of 
any standard DC, ML, or individual facility license application under 
part 53 that references the standard design approval. The proposed rule 
would also govern requirements for issuing information requests.

Sec.  53.1230 Standard Design Certifications

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

Sec.  53.1233 Filing of Applications

    This proposed rule would add Sec.  53.1233, which would enable an 
application for DC to be filed, regardless of whether an application 
for a CP, COL, or ML has been filed, provided it complies with the 
filing requirements in Sec.  53.040 and Sec. Sec.  2.811 through 2.819.

Sec.  53.1236 Contents of Applications for Standard Design 
Certifications; General Information

    This proposed rule would add Sec.  53.1236, which would require an 
application for a standard DC under part 53 to contain all of the 
information required by Sec.  53.1109(a) through (c) and (j).

Sec.  53.1239 Contents of Applications for Standard Design 
Certifications; Technical Information

    This proposed rule would add Sec.  53.1239, which would require 
applicants for a standard DC under part 53 to submit a FSAR that 
includes technical design information at a level of detail sufficient 
to enable the Commission to make a safety determination.

Sec.  53.1241 Contents of Applications for Standard Design 
Certifications; Other Application Content

    This proposed rule would add Sec.  53.1241, which would require 
applications for standard DCs under part 53 to include an environmental 
report, as well as a description of the availability controls used to 
satisfy the safety criteria of Sec.  53.220, proposed ITAAC, the 
program to protect Safeguards Information against unauthorized 
disclosure, evidence that safety questions associated with SSCs have 
been resolved, and a description of how design features fulfill design 
criteria.

Sec.  53.1242 Review of Applications

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

[[Page 87000]]

Sec.  53.1245 Referral to the Advisory Committee on Reactor Safeguards

    This proposed rule would add Sec.  53.1245, which would require the 
ACRS to report on any portions of the application for a standard DC 
under part 53 concerning safety.

Sec.  53.1248 Issuance of Standard Design Certification

    This proposed rule would add Sec.  53.1248, which would establish 
the conditions under which the Commission may issue a DC rule that 
specifies the site parameters, design characteristics, and any 
additional terms and conditions of the DC rule.

Sec.  53.1251 Duration of Certification

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

Sec.  53.1254 Application for Renewal

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

Sec.  53.1257 Criteria for Renewal

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

Sec.  53.1260 Duration of Renewal

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

Sec.  53.1263 Finality of Standard Design Certifications

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

Sec.  53.1270 Manufacturing Licenses

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

Sec.  53.1273 Filing of Applications

    This proposed rule would add Sec.  53.1273, which would establish 
the requirements to apply for an ML under part 53.

Sec.  53.1276 Contents of Applications for Manufacturing Licenses; 
General Information

    This proposed rule would add Sec.  53.1276, which would require 
applicants for an ML under part 53 to include the information contained 
in Sec.  53.1109(a) through (e) and (j).

Sec.  53.1279 Contents of Applications for Manufacturing Licenses; 
Technical Information

    This proposed rule would add Sec.  53.1279, which would require an 
applicant for an ML under part 53 to include certain technical 
information in a FSAR, including but not limited to information about 
site parameters, design information, manufacturing information, and 
information related to the potential fueling and ultimate deployment of 
a completed manufactured reactor.

Sec.  53.1282 Contents of Applications for Manufacturing Licenses; 
Other Application Content

    This proposed rule would add Sec.  53.1282, which would require 
applicants for an ML under part 53 to include in their application the 
proposed ITAAC, an environmental report, a description of the program 
to protect Safeguards Information against unauthorized disclosure, and 
a description of how design features fulfill design criteria. It would 
also include content requirements for the ITAAC and environmental 
reports in applications that reference a standard DC.

Sec.  53.1285 Review of Applications

    This proposed rule would add Sec.  53.1285, which would require 
applications for MLs under part 53 to be reviewed for compliance with 
applicable standards and establish procedural requirements for 
reviewing applicants and holding hearings in accordance with part 2.

Sec.  53.1286 Referral to the Advisory Committee on Reactor Safeguards

    This proposed rule would add Sec.  53.1286, which would require the 
ACRS to report on any portions of the application for an ML under part 
53 concerning safety.

Sec.  53.1287 Issuance of Manufacturing Licenses

    This proposed rule would add Sec.  53.1287, which would establish 
the conditions under which the Commission may issue an ML under part 
53.

Sec.  53.1288 Finality of Manufacturing Licenses

    This proposed rule would add Sec.  53.1288, which would address the 
limited circumstances in which the Commission may modify, rescind, or 
impose new requirements following the issuance of an ML under part 53. 
It would also address requests for a departure from the specifications 
of the license.

Sec.  53.1291 Duration of Manufacturing Licenses

    This proposed rule would add Sec.  53.1291, which would govern the 
expiration of an ML, which is valid for no less than 5, nor more than 
15 years from the date of issuance.

Sec.  53.1293 Transfer of Manufacturing Licenses

    This proposed rule would add Sec.  53.1293, which would provide 
that an ML under part 53 may be transferred in accordance with Sec.  
53.1570.

Sec.  53.1295 Renewal of Manufacturing Licenses

    This proposed rule would add Sec.  53.1295, which would establish 
the procedures for applicants to apply for and the Commission to grant 
a renewal of an ML under part 53.

Sec.  53.1300 Construction Permits

    This proposed rule would add Sec.  53.1300, which would provide an 
overview of the requirements and procedures for applicants to apply for 
and the Commission to grant a CP under part 53.

Sec.  53.1306 Contents of Applications for Construction Permits; 
General Information

    This proposed rule would add Sec.  53.1306, which would require 
applicants for a CP under part 53 to submit the general information 
required by Sec.  53.1109, as well as financial information.

Sec.  53.1309 Contents of Applications for Construction Permits; 
Technical Information

    This proposed rule would add Sec.  53.1309, which would require 
applicants for a CP under part 53 to submit a PSAR and a description of 
the program to protect Safeguards

[[Page 87001]]

Information from unauthorized disclosure.

Sec.  53.1312 Contents of Applications for Construction Permits; Other 
Application Content

    This proposed rule would add Sec.  53.1312, which would require 
applicants for a CP under part 53 to submit an environmental report and 
to provide additional details in the PSAR if the application references 
an ESP, standard design approval, or standard DC.

Sec.  53.1315 Review of Applications

    This proposed rule would add Sec.  53.1315, which would require 
applications for CPs under part 53 to be reviewed for compliance with 
applicable standards and establish procedural requirements for 
reviewing applications and holding hearings in accordance with part 2.

Sec.  53.1318 Finality of Referenced NRC Approvals, Permits, and 
Certifications

    This proposed rule would add Sec.  53.1318, which would address the 
finality of ESPs, standard design approvals, and standard DCs 
referenced in the CP application.

Sec.  53.1324 Referral to the Advisory Committee on Reactor Safeguards

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

Sec.  53.1327 Authorization To Conduct Limited Work Authorization 
Activities

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

Sec.  53.1330 Exemptions, Departures, and Variances

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

Sec.  53.1333 Issuance of Construction Permits

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

Sec.  53.1336 Finality of Construction Permits

    This proposed rule would add Sec.  53.1336, which would address the 
finality of CPs.

Sec.  53.1342 Duration of Construction Permits

    This proposed rule would add Sec.  53.1342, which would establish 
requirements for the expiration of a CP.

Sec.  53.1345 Transfer of Construction Permits

    This proposed rule would add Sec.  53.1345, which would govern the 
transfer of CPs under part 53.

Sec.  53.1348 Termination of Construction Permits

    This proposed rule would add Sec.  53.1348, which would require the 
holder of a permit under part 53 to provide written certification to 
the Commission within 30 days of determining to permanently cease 
construction.

Sec.  53.1360 Operating Licenses

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

Sec.  53.1366 Contents of Applications for Operating Licenses; General 
Information

    This proposed rule would add Sec.  53.1366, which would require an 
application for an OL under part 53 to include the information required 
by Sec.  53.1109 as well as financial information.

Sec.  53.1369 Contents of Applications for Operating Licenses; 
Technical Information

    This proposed rule would add Sec.  53.1369, which would require an 
application for an OL under part 53 to include certain technical 
information in an FSAR at a level of detail sufficient for the 
Commission to reach a final conclusion on all safety matters.

Sec.  53.1372 Contents of Applications for Operating Licenses; Other 
Application Content

    This proposed rule would add Sec.  53.1372, which would require an 
application for an OL under part 53 to include an environmental report 
and a description of availability controls.

Sec.  53.1375 Review of Applications

    This proposed rule would add Sec.  53.1375, which would establish 
the standards and procedures for reviewing applications and holding 
hearings on OLs under part 53.

Sec.  53.1381 Referral to the Advisory Committee on Reactor Safeguards

    This proposed rule would add Sec.  53.1381, which would require the 
ACRS to report on any portions of the application for a CP under part 
53 concerning safety.

Sec.  53.1384 Exemptions, Departures, and Variances

    This proposed rule would add Sec.  53.1384, which would govern 
requests for and the issuance of exemptions from the Commission's 
regulations and exemptions, departures, and variances from NRC 
approvals, permits, and certifications.

Sec.  53.1387 Issuance of Operating Licenses

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

Sec.  53.1390 Backfitting of Operating Licenses

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

Sec.  53.1396 Duration of Operating Licenses

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

Sec.  53.1399 Transfer of an Operating License

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

Sec.  53.1402 Application for Renewal

    This proposed rule would add Sec.  53.1402, which would provide 
that an application for a renewed OL under part 53 must be filed in 
accordance with Sec.  53.1595.

Sec.  53.1405 Continuation of an Operating License

    This proposed rule would add Sec.  53.1405, which would govern the 
continuing obligations of the holder of an OL under part 53 following 
the permanent cessation of operations.

[[Page 87002]]

Sec.  53.1410 Combined Licenses

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

Sec.  53.1413 Contents of Applications for Combined Licenses; General 
Information

    This proposed rule would add Sec.  53.1413, which would require an 
application for a COL under part 53 to include the information required 
by Sec.  53.1109 as well as financial information.

Sec.  53.1416 Contents of Applications for Combined Licenses; Technical 
Information

    This proposed rule would add Sec.  53.1416, which would require 
applicants for a COL under part 53 to submit an FSAR with a level of 
technical information sufficient to reach a final conclusion on all 
safety matters.

Sec.  53.1419 Contents of Applications for Combined Licenses; Other 
Application Content

    This proposed rule would add Sec.  53.1419, which would require 
applicants for a COL under part 53 to submit an environmental report, a 
description of availability controls, the ITAAC that the licensee must 
perform. It would also include ITAAC requirements for applications that 
reference an ESP, standard DC, ML, or combination thereof.

Sec.  53.1422 Review of Applications

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

Sec.  53.1425 Finality of Referenced NRC Approvals

    This proposed rule would add Sec.  53.1425 which would address the 
finality of ESPs, standard DC rules, standard design approvals, or MLs 
referenced in the application for a COL under part 53.

Sec.  53.1431 Referral to the Advisory Committee on Reactor Safeguards

    This proposed rule would add Sec.  53.1431, which would require the 
ACRS to report on any portions of the application for a COL under part 
53 concerning safety.

Sec.  53.1434 Authorization To Conduct Limited Work Authorization 
Activities

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

Sec.  53.1437 Exemptions, Departures, and Variances

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

Sec.  53.1440 Issuance of Combined Licenses

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

Sec.  53.1443 Finality of Combined Licenses

    This proposed rule would add Sec.  53.1443, which would govern 
permissible modifications or amendments that the Commission may make to 
a COL, as well as permissible changes that a licensee may make to 
facilities and procedures as described in the FSAR.

Sec.  53.1449 Inspection During Construction

    This proposed rule would add Sec.  53.1449, which would establish 
requirements related to inspections, tests, or analyses for the holder 
of a COL under part 53.

Sec.  53.1452 Operation Under a Combined License

    This proposed rule would add Sec.  53.1452, which would establish 
requirements describing the notifications, hearings, and findings to be 
made prior to commencing facility operations.

Sec.  53.1455 Duration of a Combined License

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

Sec.  53.1456 Transfer of a Combined License

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

Sec.  53.1458 Application for Renewal

    This proposed rule would add Sec.  53.1458, which would provide 
that an application for renewal of a COL must be filed in accordance 
with Sec.  53.1595.

Sec.  53.1461 Continuation of Combined License

    This proposed rule would add Sec.  53.1461, which would govern the 
continuing obligations of the holder of a COL under part 53 following 
the permanent cessation of operations.

Sec.  53.1470 Standardization of Commercial Nuclear Plant Designs: 
Licenses To Construct and Operate Nuclear Power Reactors of Identical 
Design at Multiple Sites

    This proposed rule would add Sec.  53.1470, which would govern the 
requirements and procedures for filing and issuing applications for a 
CP, OL, or COL under part 53 in which the applicant seeks approval of 
the same design for multiple sites.

Subpart I--Maintaining and Revising Licensing-Basis Information

    This proposed rule would add subpart I, which would address the 
maintenance of licensing-basis information for part 53.

Sec.  53.1500 Licensing-Basis Information

    This proposed rule would add Sec.  53.1500, describing the purpose 
of subpart I, which would be to provide the requirements for the 
maintenance of licensing-basis information for commercial nuclear 
plants licensed under part 53.

Sec.  53.1502 Specific Terms and Conditions of Licenses

    This proposed rule would add Sec.  53.1502, which would outline the 
specific terms and conditions for obtaining a license under part 53.

Sec.  53.1505 Changes to Licensing-Basis Information Requiring Prior 
NRC Approval

    This proposed rule would add Sec.  53.1505, which would provide an 
overview of the process for licensees to request, and the Commission to 
issue, amendments to licensing-basis information under part 53.

Sec.  53.1510 Application for Amendment of License

    This proposed rule would add Sec.  53.1510, which would require 
licensees under part 53 to file an application to request an amendment 
to the license. Applicants must assess how their requested changes 
would impact the safety criteria and analysis

[[Page 87003]]

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

Sec.  53.1515 Public Notices; State Consultation

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

Sec.  53.1520 Issuance of Amendment

    This proposed rule would add Sec.  53.1520, which would outline 
criteria for the Commission to consider in issuing license amendments 
under part 53.

Sec.  53.1525 Revising Certification Information Within a Design 
Certification Rule

    This proposed rule would add Sec.  53.1525, which would address the 
requirements for applicants to request, and the Commission to grant, an 
exemption to a DC rule under part 53.

Sec.  53.1530 Revising Design Information Within a Manufacturing 
License

    This proposed rule would add Sec.  53.1530, which would require the 
holder of an ML to request an amendment under Sec.  53.1510 and, as 
applicable, Sec.  53.1520 to make changes to the design of a 
manufactured reactor. It would also outline the requirements for 
holders of a COL under part 53 to request amendments for changes to the 
design information of a manufactured reactor.

Sec.  53.1535 Amendments During Construction

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

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

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

Sec.  53.1545 Updating Final Safety Analysis Reports

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

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

    This proposed rule would add Sec.  53.1550, which would require 
licensees under part 53 to follow the guidelines outlined in this 
section in determining whether changes to licensing-basis information 
described in the FSAR (as updated) require them to obtain a license 
amendment.

Sec.  53.1560 Updating Program Documents Included in Licensing-Basis 
Information

    This proposed rule would add Sec.  53.1560, which would require the 
holders of an OL or COL under part 53 to regularly update the program 
documents that they submitted in their application for a license.

Sec.  53.1565 Evaluating Changes to Programs Included in Licensing-
Basis Information

    This proposed rule would add Sec.  53.1565, which would enable 
licensees under part 53 to make changes to the facility, procedures, or 
organization, or address changes to site environs as described in 
program documents without NRC approval if these changes satisfy the 
criteria outlined in this section.

Sec.  53.1570 Transfer of Licenses

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

Sec.  53.1575 Termination of Licenses

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

Sec.  53.1580 Information Requests

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

Sec.  53.1585 Revocation, Suspension, Modification of Licenses and 
Approvals for Cause

    This proposed rule would add Sec.  53.1585, which would address 
grounds for the revocation, suspension, or modification of a license or 
standard design approval issued under part 53.

Sec.  53.1590 Backfitting

    This proposed rule would add Sec.  53.1590, which would define 
backfitting and establish requirements to be met by the NRC when it 
takes backfitting actions under part 53.

Sec.  53.1595 Renewal

    This proposed rule would add Sec.  53.1595, which would provide for 
the renewal of a license under part 53 upon expiration.

Subpart J--Reporting and Other Administrative Requirements

    This proposed rule would add subpart J, to establish various 
reporting and other administrative requirements for licensees under 
part 53.

Sec.  53.1600 General Information

    This proposed rule would add Sec.  53.1600, which provides an 
overview of the sections that would require applicants and licensees 
under part 53 to provide NRC inspectors with unfettered access to sites 
and facilities, maintain records and make reports, demonstrate 
compliance with financial qualification and reporting requirements, and 
maintain required financial protection for accidents.

Sec.  53.1610 Unfettered Access for Inspections

    This proposed rule would add Sec.  53.1610, which would require 
applicants and licensees under part 53 to provide unfettered access to 
NRC inspectors, including access to records, premises, activities, and 
licensed materials, in addition to providing office space to 
accommodate temporary or resident inspectors.

Sec.  53.1620 Maintenance of Records, Making of Reports

    This proposed rule would add Sec.  53.1620, which would require 
part 53 licensees to maintain all records and make reports as required 
by the conditions of the license or by the regulations in part 53.

Sec.  53.1630 Immediate Notification Requirements for Operating 
Commercial Nuclear Plants

    This proposed rule would add Sec.  53.1630, which would impose 
immediate notification requirements on

[[Page 87004]]

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

Sec.  53.1640 Licensee Event Report System

    This proposed rule would add Sec.  53.1640, which would require any 
commercial plant licensee holding an OL under part 53 to submit a 
Licensee Event Report in accordance with the specifications outlined in 
this section.

Sec.  53.1645 Reports of Radiation Exposure to Members of the Public

    The proposed rule would add Sec.  53.1645, which would require 
annual reports to the Commission, including radiological reports as 
required by part 20, an Annual Radioactive Effluent Release Report, and 
an Annual Environmental Operating Report.

Sec.  53.1650 Facility Information and Verification

    The proposed rule would add Sec.  53.1650, which would include a 
reporting requirement for applicants and holders of a CP or license 
under part 53 to support safeguards agreements between the United 
States and the IAEA.

Sec.  53.1660 Financial Requirements

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

Sec.  53.1670 Financial Qualifications

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

Sec.  53.1680 Annual Financial Reports

    This proposed rule would add Sec.  53.1680, which would require 
licensees and holders of a CP under part 53 to submit annual financial 
reports to the Commission, with exceptions for those that submit 
financial forms to the Securities and Exchange Commission or the 
Federal Energy Regulatory Commission.

Sec.  53.1690 Licensee's Change of Status; Financial Qualifications

    This proposed rule would add Sec.  53.1690, which would require 
electric utility licensees that hold an OL or COL for a commercial 
nuclear plant under part 53 to provide the NRC with the financial 
qualifications information outlined in this section within seventy-five 
days of ceasing to be an electric utility.

Sec.  53.1700 Creditor Regulations

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

Sec.  53.1710 Financial Protection

    This proposed rule would add Sec.  53.1710, which would establish 
requirements for licenses under part 53 to obtain and maintain 
insurance to cover the costs of an accident.

Sec.  53.1720 Insurance Required To Stabilize and Decontaminate Plant 
Following an Accident

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

Sec.  53.1730 Financial Protection Requirements

    This proposed rule would add Sec.  53.1730, which would require 
commercial nuclear plant licensees under part 53 to satisfy the 
provisions of part 140.

Subpart M--Enforcement

    This proposed rule would add subpart M, which would address certain 
violations and penalties associated with violations of part 53 
regulations.

Sec.  53.9000 Violations

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

Sec.  53.9010 Criminal Penalties

    This proposed rule would add Sec.  53.9010, providing notice to all 
persons and entities subject to part 53 that they are subject to 
criminal sanctions for willful violations, attempted violations, or 
conspiracy to violate certain regulations under part 53.

Sec.  70.20a General License to Possess Special Nuclear Material for 
Transport

    This proposed rule would revise Sec.  70.20a(b) to include a 
reference to part 53.

Sec.  70.22 Contents of Applications

    This proposed rule would revise Sec.  70.22, paragraphs (b), 
(h)(1), (j)(1), and (k) to include the appropriate references to part 
53.

Sec.  70.24 Criticality Accident Requirements

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

Sec.  70.32 Conditions of Licenses

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

Sec.  70.50 Reporting Requirements

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

Sec.  72.3 Definitions

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

Sec.  72.30 Financial Assurance and Recordkeeping for Decommissioning

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

Sec.  72.32 Emergency Plan

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

Sec.  72.40 Issuance of License

    This proposed rule would revise Sec.  72.40(c) regarding the 
issuance of a license under part 72 to include a reference to previous 
licensing actions, including the issuance of a CP under part 53.

Sec.  72.75 Reporting Requirements for Specific Events and Conditions

    This proposed rule would revise Sec.  72.75(i)(1)(ii) regarding 
reporting requirements for specific events and conditions with 
references to reactors licensed under part 53.

Sec.  72.184 Safeguards Contingency Plan

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

Sec.  72.210 General License Issued

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

[[Page 87005]]

Sec.  72.212 Conditions of General License Issued Under Sec.  72.210

    This proposed rule would revise Sec.  72.212(b)(8) regarding the 
conditions of a general license issued under Sec.  72.210 to include a 
reference to license amendments for a facility made pursuant to part 
53.

Sec.  72.218 Termination of Licenses

    This proposed rule would revise Sec.  72.218(a) to include a 
reference to the notification required under part 53 regarding the plan 
for managing spent fuel prior to decommissioning. It would also extend 
the provisions of Sec.  72.218(b) to a reactor operating or COL under 
part 53.

Sec.  73.1 Purpose and Scope

    This proposed rule would revise Sec.  73.1(b)(1)(i) to extend the 
scope of part 73 to production and utilization facilities licensed 
under part 53, in addition to parts 50 and 52.

Sec.  73.2 Definitions

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

Sec.  73.8 Information Collection Requirements: OMB Approval

    This proposed rule would revise Sec.  73.8(b) with the new 
information collection requirements contained in proposed Sec. Sec.  
73.77, 73.100, 73.110, and 73.120.

Sec.  73.50 Requirements for Physical Protection of Licensed Activities

    This proposed rule would revise Sec.  73.50 to exempt nuclear 
reactor facilities licensed under part 53, in addition to parts 50 and 
52, from the requirements of this section.

Sec.  73.55 Requirements for Physical Protection of Licensed Activities 
in Nuclear Power Reactors Against Radiological Sabotage

    This proposed rule would revise Sec.  73.55, paragraphs (a)(4) and 
(6), (i)(4)(iii), (l)(1), (l)(7)(ii), (p)(1)(i), (r)(2), and 
(r)(4)(iii), to incorporate the appropriate references to part 53 
regarding requirements for physical protection of licensed activities 
in nuclear power reactors against radiological sabotage.

Sec.  73.56 Personnel Access Authorization Requirements for Nuclear 
Power Plants

    This proposed rule would revise Sec.  73.56(a)(3) to apply this 
section's personnel AA requirements to applicants for an OL or holders 
of a COL under part 53 who do not demonstrate compliance with certain 
requirements under part 53.

Sec.  73.57 Requirements for Criminal History Records Checks of 
Individuals Granted Unescorted Access to a Nuclear Power Facility, a 
Non-power Reactor, or Access to Safeguards Information

    This proposed rule would revise Sec.  73.57(a)(3) to incorporate 
the appropriate references to OLs granted under part 53 and Commission 
findings under Sec.  53.1452(g) regarding the requirement for license 
applicants to submit fingerprints for all personnel with unescorted 
access.

Sec.  73.58 Safety/Security Interface Requirements for Nuclear Power 
Reactors

    This proposed rule would revise Sec.  73.58(a) to extend the 
requirements of this section to part 53 licensees.

Sec.  73.67 Licensee Fixed Site and In-Transit Requirements for the 
Physical Protection of Special Nuclear Material of Moderate and Low 
Strategic Significance

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

Sec.  73.77 Cybersecurity Event Notifications

    This proposed rule would revise Sec.  73.77, paragraphs (a), (b), 
(c)(6) and (7) regarding the notification process for cybersecurity 
events to include notifications for the declaration of an emergency 
class made under part 53.

Subpart J--Security Requirements at Commercial Nuclear Plants

    This proposed rule would add new Subpart J of part 73 containing 
Sec. Sec.  73.100, 73.110, and 73.120, to establish security 
requirements for commercial nuclear plants licensed under part 53.

Sec.  73.100 Technology-Inclusive Requirements for Physical Protection 
of Licensed Activities at Commercial Nuclear Plants Against 
Radiological Sabotage

    This proposed rule would add Sec.  73.100, which would establish a 
performance-based regulatory framework for physical protection as an 
alternative to the prescriptive requirements of Sec.  73.55, which also 
governs physical protection programs for part 50 and 52 licensees.

Sec.  73.110 Technology-Inclusive Requirements for Protection of 
Digital Computer and Communication Systems and Networks

    This proposed rule would add Sec.  73.110, which would establish a 
consequence-based approach to cybersecurity and would require that part 
53 licensees demonstrate reasonable assurance that digital computer and 
communication systems and networks are adequately protected against 
cyberattacks in a manner that is commensurate with the potential 
consequences of those attacks.

Sec.  73.120 Access Authorization Program for Commercial Nuclear Plants

    This proposed rule would add Sec.  73.120, which would establish 
performance objectives as an alternative to compliance with the AA 
provisions of Sec. Sec.  73.55, 73.56, and 73.57. This proposed rule 
would afford part 53 licensees additional flexibility in establishing 
an AA program that demonstrates compliance with the performance 
objectives and requirements of this section.

Sec.  73.1200 Notification of Physical Security Events

    This proposed rule would revise Sec.  73.1200, paragraphs (a), 
(c)(1), (e)(1), (e)(3), (e)(4), (g)(1), (o)(5)(i), (o)(6)(i), (r), and 
(s) to extend the requirements of this section to part 53 licensees.

Sec.  73.1205 Written Follow-Up Reports of Physical Security Events

    This proposed rule would revise Sec.  73.1205(b)(2) to extend the 
requirements of this section to part 53 licensees.

Sec.  73.1210 Recordkeeping of Physical Security Events

    This proposed rule would revise Sec.  73.1210(a)(1) and (b)(3)(i) 
to extend the requirements of this section to part 53 licensees.

Sec.  73.1215 Suspicious Activity Reports

    This proposed rule would revise Sec.  73.1215(d)(1) to include a 
reference to Sec.  73.100.

Appendix B to part 73--General Criteria for Security Personnel

    This proposed rule would revise appendix B to part 73 to state that 
terms defined in part 53 have the same meaning when used in this 
appendix.

Sec.  74.31 Nuclear Material Control and Accounting for Special Nuclear 
Material of Low Strategic Significance

    This proposed rule would revise Sec.  74.31(a) to include a 
reference to

[[Page 87006]]

production or utilization facilities licensed under part 53, in 
addition to parts 50 and 70.

Sec.  74.41 Nuclear Material Control and Accounting for Special Nuclear 
Material of Moderate Strategic Significance

    This proposed rule would revise Sec.  74.41(a) to include a 
reference to nuclear reactors licensed under part 53.

Sec.  74.51 Nuclear Material Control and Accounting for Strategic 
Special Nuclear Material

    This proposed rule would revise Sec.  74.51(a) to include a 
reference to nuclear reactors licensed under part 53.

Sec.  75.4 Definitions

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

Sec.  95.5 Definitions

    This proposed rule would revise the definition of ``License'' in 
Sec.  95.5 to include those issued under part 53.

Sec.  95.39 External Transmission of Documents and Material

    This proposed rule would revise Sec.  95.39(a) to apply 
restrictions to the external transmission of documents and material 
containing classified information in connection with NRC licenses, 
certificates, standard design approvals, or standard DCs issued under 
part 53.

Sec.  140.2 Scope

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

Sec.  140.10 Scope

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

Sec.  140.11 Amounts of Financial Protection for Certain Reactors

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

Sec.  140.12 Amount of Financial Protection Required for Other Reactors

    This proposed rule would revise Sec.  140.12(c) to require the 
licensee's primary financial protection to cover all reactors in any 
case where a person is authorized under part 53 to operate two or more 
nuclear reactors at the same location.

Sec.  140.13 Amount of Financial Protection Required of Certain Holders 
of Construction Permits and Combined Licenses Under 10 CFR Part 52

    This proposed rule would revise Sec.  140.13 with the appropriate 
references to part 53 regarding the requirement for holders of a CP or 
COL under part 53 to obtain financial protection.

Sec.  140.20 Indemnity Agreements and Liens

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

Sec.  150.15 Persons Not Exempt

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

Sec.  170.3 Definitions

    The proposed rule would revise Sec.  170.3 to incorporate 
references to part 53 into the definitions of ``Manufacturing 
license,'' ``Part 55 Reviews,'' ``Power reactor,'' and ``Special 
projects.''

Sec.  170.12 Payment of Fees

    The proposed rule would revise Sec.  170.12(d)(1)(v) regarding 
special project fees in connection with FSARs to include part 53.

Sec.  170.21 Schedule of Fees for Production and Utilization 
Facilities, Review of Standard Referenced Design Approvals, Special 
Projects, Inspections, And import and Export Licenses

    The proposed rule would revise Sec.  170.21, footnote 1 to include 
fees charged for approvals issued under the exemption provision in 
Sec.  53.080.

Sec.  170.41 Failure by Applicant or Licensee to Pay Prescribed Fees

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

Sec.  171.3 Scope

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

Sec.  171.5 Definitions

    This proposed rule would revise the definitions of ``Operating 
license'' and ``Power reactor'' in Sec.  171.5 to incorporate the 
appropriate references to part 53.

Sec.  171.15 Annual fees: Non-Power Production or Utilization Licenses, 
Reactor Licenses, and Independent Spent Fuel Storage Licenses

    This proposed rule would revise Sec.  171.15, paragraphs (a), 
(b)(2)(iii), (c)(1), and (d)(1) regarding annual fees that are 
applicable to part 53 licensees.

Sec.  171.17 Proration

    This proposed rule would revise Sec.  171.17, paragraphs (a), 
(a)(1)(ii) and (a)(2) with references to part 53 licenses.

VIII. Regulatory Flexibility Certification

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

[[Page 87007]]

owned and operated and has annual gross receipts of $8.0 million or 
less.
    Small governmental jurisdiction is a government of a city, county, 
town, township, village, school district, or special district with a 
population of less than 50,000.
    Small educational institution is one that is--(1) Supported by a 
qualifying small governmental jurisdiction; or (2) Not State or 
publicly supported and has 500 or fewer employees.

Number of Small Entities Affected

    The NRC is currently not aware of any known small entities as 
defined in Sec.  2.810 that are planning to apply for a commercial 
nuclear plant ESP, CP, OL, ML, or COL under part 53 that would be 
impacted by this proposed rule. Based on this finding, the NRC has 
preliminarily determined that the proposed rule would not have a 
significant economic impact on a substantial number of small entities.

Economic Impact on Small Entities

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

Request for Comments

    The NRC is seeking comment on both its initial RFA analysis and on 
its preliminary conclusion that this proposed rule would not have a 
significant economic impact on a substantial number of small entities 
because of the likelihood that most expected applicants would not 
qualify as a small entity. Additionally, the NRC is seeking comment on 
its preliminary conclusion that if a small entity were to submit a 
commercial nuclear plant application, the small entity would not incur 
a significant economic impact as it would most likely not be in 
competition with a large entity.
    Any small entity that could be subject to this regulation that 
determines, because of its size, it is likely to bear a 
disproportionate adverse economic impact should notify the Commission 
of this opinion in a comment that indicates--
    1. The applicant's size and how the proposed regulation would 
impose a significant economic burden on the applicant as compared to 
the economic burden on a larger applicant;
    2. How the proposed regulations could be modified to take into 
account the applicant's differing needs or capabilities;
    3. The benefits that would accrue or the detriments that would be 
avoided if the proposed regulations were modified as suggested by the 
applicant;
    4. How the proposed regulation, as modified, would more closely 
equalize the impact of NRC regulations or create more equal access to 
the benefits of Federal programs as opposed to providing special 
advantages to any individual or group; and
    5. How the proposed regulation, as modified, would still adequately 
demonstrate compliance with the NRC's obligations under the Act.

IX. Regulatory Analysis

    The NRC has prepared a draft regulatory analysis for this proposed 
rule. The analysis examines the costs and benefits of the alternatives 
considered by the NRC. The conclusion from the analysis is that this 
proposed rule and associated guidance would result in net averted costs 
to the industry and the NRC of $28.1 million using a 7-percent discount 
rate and $34.5 million using a 3-percent discount rate due to 
reductions in exemption requests. The analysis also assumes one 
applicant under part 53. As the number of applicants increases, so do 
the estimated averted costs. The NRC requests public comment on the 
draft regulatory analysis, which is available as indicated in the 
``Availability of Documents'' section of this document. Comments on the 
draft regulatory analysis may be submitted to the NRC as indicated 
under the ADDRESSES caption of this document.

X. Backfitting and Issue Finality

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

[[Page 87008]]

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

XI. Cumulative Effects of Regulation

    The NRC seeks to minimize any potential negative consequences 
resulting from the cumulative effects of regulation (CER). The CER 
describes the challenges that licensees, or other impacted entities 
such as State partners, may face while implementing new regulatory 
positions, programs, or requirements (e.g., rules, generic letters, 
backfits, inspections). The CER is an organizational effectiveness 
challenge that may result from a licensee or impacted entity 
implementing a number of complex regulatory actions, programs, or 
requirements within limited available resources. The NRC's CER process 
involved engaging with external stakeholders throughout this proposed 
rule and related regulatory activities. Public involvement has included 
numerous public meetings to examine the part 53 risk-informed, 
technology-inclusive requirements for commercial nuclear plants and the 
publication of numerous versions of preliminary proposed rule language. 
The NRC is considering holding additional public meetings during the 
remainder of the rulemaking process.
    In parallel with this proposed rule, the NRC is issuing 10 draft 
implementing guidance documents for comment to support informed 
external stakeholder feedback. Section XVII, ``Availability of 
Guidance,'' of this document describes how the public can access the 
draft implementing guidance.
    In addition to the questions in the ``Specific Requests for 
Comments'' section of this document, the NRC is requesting CER feedback 
on the following questions:
    1. In light of any current or projected CER challenges, does the 
proposed rule's effective date provide sufficient time to implement the 
new proposed requirements, including changes to programs, procedures, 
and the facility?
    2. If CER challenges currently exist or are expected, what should 
be done to address them? For example, if more time is required for 
implementation of the new requirements, what period of time is 
sufficient?
    3. Do other (NRC or other agency) regulatory actions (e.g., orders, 
generic communications, license amendment requests, inspection findings 
of a generic nature) influence the implementation of the proposed 
rule's requirements?
    4. Are there unintended consequences? Does the proposed rule create 
conditions that would be contrary to the proposed rule's purpose and 
objectives? If so, what are the unintended consequences, and how should 
they be addressed?
    5. Please comment on the NRC's cost and benefit estimates in the 
regulatory analysis that supports this proposed rule. The draft 
regulatory analysis is available as indicated under the ``Availability 
of Documents'' section of this document.

XII. Plain Writing

    The Plain Writing Act of 2010 (Pub. L. 111-274) requires Federal 
agencies to write documents in a clear, concise, and well-organized 
manner. The NRC has written this document to be consistent with the 
Plain Writing Act as well as the Presidential Memorandum, ``Plain 
Language in Government Writing,'' published June 10, 1998 (63 FR 
31885). The NRC requests comment on this document with respect to the 
clarity and effectiveness of the language used.

XIII. Environmental Assessment and Proposed Finding of No Significant 
Environmental Impact

    The Commission has preliminarily determined under the National 
Environmental Policy Act of 1969, as amended, and the Commission's 
regulations in subpart A of part 51, that this rule, if adopted, would 
not be a major Federal action significantly affecting the quality of 
the human environment, and an EIS is not required. The implementation 
of the proposed rule requirements does not have a significant impact on 
the environment. The proposed rulemaking would either have requirements 
that are administrative in application, matters of procedure, or 
provide an equivalent level of safety as existing requirements; 
therefore, there would be similar environmental impacts from the 
implementation of the part 53 regulations as there are for existing 
requirements.
    The preliminary determination of this EA is that there will be no 
significant effect on the quality of the human environment from this 
action. Public stakeholders should note, however, that comments on any 
aspect of this EA may be submitted to the NRC as indicated under the 
ADDRESSES section of this document. The EA is available as indicated 
under the ``Availability of Documents'' section of this document.
    The NRC has sent a copy of the EA, and this proposed rule to every 
State Liaison Officer and has requested comments.

XIV. Paperwork Reduction Act

    This proposed rule contains new collections of information 
contained in parts 26, 50, 53, and 73 and NRC Forms 361S, 366, 366A, 
366B, 893, and 894 that are subject to the Paperwork Reduction Act of 
1995 (44 U.S.C. 3501 et seq). The collections of information have been 
submitted to the OMB for review and approval. The proposed changes to 
parts 2, 10, 11, 19, 20, 21, 25, 30, 40, 51, 70, 72, 74, 75, 95, 140, 
150, 170, and 171 do not contain any new or amended collections of 
information subject to the Paperwork Reduction Act of 1995. Existing 
collections of information were approved by the OMB, approval numbers 
3150-0062 (part 11), 3150-0044 (part 19), 3150-0014 (part 20), 3150-
0035 (part 21), 3150-0046 (part 25), 3150-0017 (part 30), 3150-0020 
(part 40), 3150-0021 (part 51), 3150-0024 (NRC Form 396), 3150-0090 
(NRC Form 398), 3150-0009 (part 70), 3150-0132 (part 72), 3150-0123 
(part 74), 3150-0055 (part 75), 3150-0047 (part 95), 3150-0039 (part 
140), and 3150-0032 (part 150).
    Type of submission, new or revision: Revision and new.
    The title of the information collection: Risk-Informed, Technology-
Inclusive Regulatory Framework for Advanced Reactors.

[[Page 87009]]

    The form number if applicable: NRC Forms 361S, 366, 366A, 366B, 
893, and 894.
    How often the collection is required or requested: Once, on 
occasion, every 30 days, biannually, annually, biennially, every four 
years, every five years, every ten years.
    Who will be required or asked to respond: Part 53 commercial 
nuclear plant licensees and license applicants for commercial nuclear 
plants to be licensed under part 53.
    An estimate of the number of annual responses: 15 (2 responses for 
Part 26, 11 responses for Part 53, 2 responses for Part 50 and 0 
responses for Part 73 and NRC Forms 361S, 366, 366A, 366B, 893, and 
894)
    The estimated number of annual respondents: 2 (2 respondents for 
Part 26, 2 respondents for Part 53, 2 respondents for Part 50 and 0 
respondents for Part 73 and NRC Forms 361S, 366, 366A, 366B, 893, and 
894)
    An estimate of the total number of hours needed annually to comply 
with the information collection requirement or request: 230,244 hours. 
(656 hours for Part 26, 220,801 hours for Part 53, 8,767 hours for Part 
50 and 0 hours for Part 73 and NRC Forms 361S, 366, 366A, 366B, 893, 
and 894)
    Abstract: The NRC is proposing to establish an optional technology-
inclusive regulatory framework for use by applicants for new commercial 
nuclear plant designs. The regulatory requirements developed in this 
rulemaking would use methods of evaluation, including risk-informed and 
performance-based methods, that are flexible and practicable for 
application to a variety of new reactor technologies. The NRC's goals 
in amending these regulations are to continue to provide reasonable 
assurance of adequate protection of public health and safety and the 
common defense and security at reactor sites at which new nuclear 
reactor designs are deployed to at least the same degree of protection 
as required for current-generation LWRs; protect health and minimize 
danger to life or property to at least the same degree of protection as 
required for current-generation LWRs; provide greater operational 
flexibilities where supported by enhanced margins of safety that may be 
provided in new nuclear designs; and promote regulatory stability, 
predictability, and clarity.
    The proposed rule covers diverse topics, which result in 
recordkeeping and reporting requirements related to contents of 
applications, plant design and analysis, siting, construction and 
manufacturing, licensing-basis information, facility operations, 
programs, staffing, FFD, physical security, cyber-security, AA, 
decommissioning, and quality assurance.
    In addition to the new information collections in the proposed 
regulations, part 53 would result in new collections via NRC Forms 
361S, 366, 366A, 366B, 893, and 894. NRC Forms 366, 366A, and 366B 
would be modified to include part 53 reportable events covering an 
equivalent scope as the requirements in 10 CFR 50.73, but without LWR-
specific terminology to ensure technology inclusiveness. The proposed 
rule also would require part 53 licensees to use NRC Forms 893 and 894 
to report on positive drug and alcohol test results (NRC Form 893) and 
annual fitness-for-duty program performance (NRC Form 894). Finally, a 
new version of NRC Form 361 (NRC Form 361S) would be created for use by 
part 53 licensees, covering an equivalent scope as the requirements in 
10 CFR 50.72, but without LWR-specific terminology to ensure technology 
inclusiveness.
    The NRC is seeking public comment on the potential impact of the 
information collections contained in this proposed rule and on the 
following issues:
    1. Is the proposed information collection necessary for the proper 
performance of the functions of the NRC, including whether the 
information will have practical utility? Please explain your response.
    2. Is the estimate of the burden of the proposed information 
collection accurate? Please explain your response.
    3. Is there a way to enhance the quality, utility, and clarity of 
the information to be collected? Please explain your response.
    4. How can the burden of the proposed information collection on 
respondents be minimized, including the use of automated collection 
techniques or other forms of information technology? Please explain 
your response.
    The OMB clearance documents and proposed rule is available as 
indicated under the ``Availability of Documents'' section in this 
document or may be viewed free of charge by contacting the NRC's PDR 
reference staff at 1-800-397-4209, at 301-415-4737, or by email to 
[email protected]. You may obtain information and comment 
submissions related to the OMB clearance package by searching on http://www.regulations.gov under Docket ID NRC-2019-0062.
    You may submit comments on any aspect of these proposed information 
collections, including suggestions for reducing the burden and on the 
above issues, by the following methods:
     Federal Rulemaking Website: Go to http://www.regulations.gov and search for Docket ID NRC-2019-0062.
     Mail comments to: FOIA, Library, and Information 
Collections Branch, Office of the Chief Information Officer, Mail Stop: 
T6-A10M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 
or by email to [email protected] or to the OMB reviewer at: 
OMB Office of Information and Regulatory Affairs (3150-XXXX, 3150-0002, 
-0104, -0146, -0238), Attn: Desk Officer for the Nuclear Regulatory 
Commission, 725 17th Street NW, Washington, DC 20503.
    Submit comments by December 2, 2024. Comments received after this 
date will be considered if it is practical to do so, but the NRC staff 
is able to ensure consideration only for comments received on or before 
this date.

Public Protection Notification

    The NRC may not conduct or sponsor, and a person is not required to 
respond to, a collection of information unless the document requesting 
or requiring the collection displays a currently valid OMB control 
number.

XV. Criminal Penalties

    For the purposes of Section 223 of the Act, the NRC is issuing this 
proposed rule that would add a new part 53 and amend parts 26 and 73 
under one or more of Sections 161b, 161i, or 161o of the Act, except as 
noted in proposed Sec.  53.9010(b) and Sec.  26.825(b). Willful 
violations of the part 53 and part 26 regulations not listed in 
proposed Sec.  53.9010(b) and Sec.  26.825(b) would be subject to 
criminal enforcement. Criminal penalties as they apply to regulations 
in part 53 would be discussed in Sec.  53.9010.

XVI. Voluntary Consensus Standards

    The NTTAA requires that Federal agencies use technical standards 
that are developed or adopted by voluntary consensus standards bodies 
unless the use of such a standard is inconsistent with applicable law 
or otherwise impractical. In this proposed rule, the NRC would revise 
regulations by adding a risk-informed, technology-inclusive regulatory 
framework for commercial advanced nuclear reactors. This action does 
not constitute the establishment of a standard that contains generally 
applicable requirements.

XVII. Availability of Guidance

    As discussed in section II, Background, of this document, the NRC's 
development of proposed part 53

[[Page 87010]]

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

 RG 1.233, ``Guidance for a Technology-Inclusive, Risk-
Informed, and Performance-Based Methodology to Inform the Licensing 
Basis and Content of Applications for Licenses, Certifications, and 
Approvals for Non-Light-Water Reactors''
 RG 1.247 for trial use, ``Acceptability of Probabilistic Risk 
Assessment Results for Non-Light-Water Reactor Risk-Informed 
Activities''
 NUREG-2246, ``Fuel Qualification for Advanced Reactors''
 RG 1.87, Revision 2, ``Acceptability of ASME Code, Section 
III, Division 5, ``High Temperature Reactors''
 RG 1.246, ``Acceptability of ASME Code, Section XI, Division 
2, `Requirements for Reliability And Integrity Management (RIM) 
Programs for Nuclear Power Plants,' for Non-Light Water Reactors''

    Also, the NRC continues to develop additional guidance to support 
licensing reviews of advanced reactors under the existing regulatory 
framework. Some of these guidance documents have been issued and others 
will be issued before the finalization of part 53 to support near-term 
applicants and NRC reviews. For example, the NRC has been and continues 
to be engaged with the DOE and industry to develop content of 
application guidance and other regulatory guidance for advanced 
reactors to support applications and subsequent operations under the 
existing regulatory framework. These guidance documents, such as the 
industry-led Technology-Inclusive Content of Application Project 
guidance found in NEI 21-07, Revision 1, and the NRC-led Advanced 
Reactor Content of Application Project (ARCAP) interim staff guidance 
(ISG) documents and NRC regulatory guidance endorsing NEI 21-07, 
Revision 1, will support developers in preparing advanced reactor 
applications. These guidance documents provide an overview of the 
information that should be included in an advanced reactor application, 
a review roadmap for the NRC with the principal purpose of ensuring 
consistency, quality, and uniformity of NRC reviews, and a well-defined 
base from which the NRC can evaluate proposed changes in the scope and 
requirements of reviews. While specific sections of the information are 
primarily aligned with the LMP methodology, as endorsed in RG 1.233, as 
one acceptable process for applicants to use when developing portions 
of an application, the concepts and general information may be used to 
inform the review of an application submitted using other traditional 
licensing approach methodologies (as applicable). Other sections of the 
information are generally applicable and independent of the methodology 
used to develop an advanced reactor application. The ARCAP ISGs provide 
references to numerous regulatory guidance documents that should be 
considered by both applicants and the NRC in developing and reviewing, 
respectively, advanced reactor applications. The NRC has issued the 
following documents separately from this proposed rule. The NRC may 
issue other, related guidance documents with or after the final part 53 
rule.

 RG 1.253, ``Guidance for a Technology Inclusive Content of 
Application Methodology to Inform the Licensing Basis and Content of 
Applications for Licenses, Certifications, and Approvals for Non-Light-
Water Reactors''
 DANU-ISG-2022-01, ``Advanced Reactor Content of Application 
Project, `Review of Risk-Informed, Technology-Inclusive Advanced 
Reactor Applications--Roadmap' ''
 DANU-ISG-2022-02, ``Advanced Reactor Content of Application 
Project Chapter 2, `Site Information' ''
 DANU-ISG-2022-03, ``Advanced Reactor Content of Application 
Project Chapter 9, `Control of Routine Plant Radioactive Effluents, 
Plant Contamination and Solid Waste' ''
 DANU-ISG-2022-04, ``Advanced Reactor Content of Application 
Project Chapter 10, `Control of Occupational Dose' ''
 DANU-ISG-2022-05, ``Advanced Reactor Content of Application 
Project Chapter 11, `Organization and Human-System Considerations' ''
 DANU-ISG-2022-06, ``Advanced Reactor Content of Application 
Project Chapter 12, `Post-Construction Inspection, Testing, and 
Analysis Program' ''
 DANU-ISG-2022-07, ``Advanced Reactor Content of Application 
Project, `Risk-Informed Inservice Inspection/Inservice Testing' ''
 DANU-ISG-2022-08, ``Advanced Reactor Content of Application 
Project, `Risk-Informed Technical Specifications' ''
 DANU-ISG-2022-09, ``Advanced Reactor Content of Application 
Project, `Risk-Informed, Performance-Based Fire Protection Program (for 
Operations)' ''
 RG 1.242, ``Performance-Based Emergency Preparedness for Small 
Modular Reactors, Non-Light-Water Reactors, and Non-Power Production or 
Utilization Facilities''
 RG 4.7, ``General Site Suitability Criteria for Nuclear Power 
Stations''


[[Page 87011]]


    (2) The NRC is issuing for comment nine draft guidance documents 
for the implementation of the proposed requirements in this rulemaking. 
The guidance is available in ADAMS under the Accession Numbers as 
indicated under the ``Availability of Documents'' section in this 
document. Comments on this draft regulatory guidance may be submitted 
by the methods outlined in the ADDRESSES section of this document. 
Interested persons may obtain information and comment submissions 
related to the draft guidance by searching on http://www.regulations.gov under Docket ID NRC-2019-0062.

 DG-1413, ``Technology-Inclusive Identification of Licensing 
Events for Commercial Nuclear Plants''

    This DG describes an acceptable approach for identifying licensing 
events that can be used to inform the design basis, licensing basis, 
and content of applications for commercial nuclear plants, including 
large LWRs and non-LWRs. It applies to nuclear power reactor designers, 
applicants, and licensees of commercial nuclear plants applying for 
permits, licenses, certifications, and approvals under parts 50, 52, 
and 53. In this DG, the term ``licensing events'' is used in a generic 
sense to refer to collections of designated event categories such as, 
but not limited to AOOs, DBAs, DBEs, and postulated accidents. 
Specifically, this DG provides an acceptable approach for: (1) 
conducting a comprehensive and systematic search for initiating events; 
(2) using a systematic process to delineate a comprehensive set of 
event sequences; (3) grouping initiating events and event sequences 
into designated licensing event categories; and (4) providing assurance 
that the set of licensing events is complete.

 DG-5073, ``Fitness For Duty Programs for Commercial Nuclear 
Plants And Manufacturing Facilities Licensed Under 10 CFR part 53''

    This DG describes guidance for applicants under part 53 and 
licensees and other entities described in Sec.  26.3(f) who would elect 
to or be required to implement FFD programs for facilities licensed 
under part 53. The FFD program requirements would be detailed in 
subpart M of part 26 and involve, in part, policies, procedures, drug 
and alcohol testing, laboratory requirements, behavioral observation, 
MRO responsibilities, fitness determinations, reporting, and 
recordkeeping. The FFD program for facilities licensed under part 53 
subject to part 26 would also include requirements for a PMRP and FFD 
program change control that licensees or other entities must implement 
to maintain an effective FFD program.

 DG-5074, ``Access Authorization Program for Commercial Nuclear 
Plants''

    This DG describes a method that the staff considers acceptable to 
comply with requirements in proposed Sec.  73.120, ``Access 
authorization program for commercial nuclear plants,'' related to an AA 
program. This document provides guidance and would be one NRC-approved 
method (not the only method) for meeting regulatory requirements for 
part 53. The proposed language in Sec.  73.120 would provide 
flexibility through availability of the use of an alternate approach, 
commensurate with risk and consequence to public health and safety, for 
part 53 applicants who demonstrate in an analysis that the offsite 
consequences satisfy the criterion defined in proposed Sec.  
53.860(a)(2)(i).

 DG-5075, ``Establishing Cybersecurity Programs for Commercial 
Nuclear Plants Licensed Under 10 CFR part 53''

    This DG describes an approach the NRC staff deems acceptable for 
complying with the Commission's proposed regulations for establishing, 
implementing, and maintaining a cybersecurity program at commercial 
nuclear plants that would be licensed under part 53. This guidance 
provides an approach for meeting the requirements of proposed Sec.  
73.110, ``Technology-inclusive requirements for protection of digital 
computer and communication systems and networks.''

 DG-5076, ``Guidance for Technology Inclusive Requirements for 
Physical Protection of Licensed Activities at Commercial Nuclear 
Plants''

    This DG describes methods and approaches that the NRC staff 
considers acceptable for meeting the proposed physical security 
requirements of part 53 and Sec.  73.100. The guidance is intended to 
provide methods and considerations for complying with Sec.  53.440(f) 
safety and security design process considerations, determining 
eligibility for meeting the performance criterion in Sec.  53.860 to 
relieve the applicant from the applicable requirements to defend 
against radiological sabotage outlined in Sec.  73.55 or Sec.  73.100, 
and (if the required analysis for eligibility is not satisfied) 
applying the physical security requirements of Sec.  73.100 as an 
alternative pathway from Sec.  73.55 for protection against 
radiological sabotage.

 DG-5078, ``Fatigue Management for Nuclear Power Plant 
Personnel at Commercial Nuclear Plants Licensed Under 10 CFR part 53''

    This DG describes proposed methods that the NRC staff considers 
acceptable for addressing certain aspects of FFD programs that would be 
established at commercial nuclear facilities licensed under part 53. 
This guidance, in conjunction with the existing RG 5.73, ``Fatigue 
Management for Nuclear Plant Personnel,'' would provide comprehensive 
guidance regarding acceptable methods for the development and 
implementation of licensee fatigue-management programs.
    The NRC is issuing for public comment the following draft ISG 
documents for the implementation of NRC staff review of applications 
under the proposed requirements in this rulemaking:

 DRO-ISG-2023-01, ``Operator Licensing Programs''

    This draft ISG provides guidance for the review of tailored 
operator licensing programs that are submitted for review consistent 
with the technical requirements of proposed Sec.  53.730(g). This 
guidance primarily addresses the review of operator licensing 
examination processes to facilitate the ability of reviewers to assess 
whether a proposed approach to the testing of licensed operators and 
trainees reflects sound assessment testing practices that are suitable 
for the screening of competent licensed operators. Additionally, this 
ISG provides further review guidance in other areas such as licensed 
operator continuing training and proficiency programs.

 DRO-ISG-2023-02, ``Interim Staff Guidance Augmenting NUREG-
1791, `Guidance for Assessing Exemption Requests from the Nuclear Power 
Plant Licensed Operator Staffing Requirements Specified in 10 CFR 
50.54(m),' for Licensing Commercial Nuclear Plants under 10 CFR part 
53''
    This draft ISG provides guidance for the review of customized 
facility operator staffing plans that are submitted for review 
consistent with the technical requirements of proposed Sec.  53.730(f). 
This ISG is structured as a companion document to the existing NUREG-
1791 and adapts the existing HFE-based methodologies of that document 
for use in the evaluation of staffing plans that would be submitted 
within the context of part 53 facilities. Additionally, this ISG 
provides further guidance to address other staffing-related 
considerations, such as provisions for engineering expertise.

 DRO-ISG-2023-03, ``Development of Scalable Human Factors 
Engineering Review Plans''


[[Page 87012]]


    This draft ISG applies to the HFE review of applications for OLs, 
COLs, DCs, and standard design approvals for commercial nuclear plants 
submitted under proposed part 53. The purpose of this ISG is to 
facilitate NRC understanding of an acceptable method for developing a 
scalable (i.e., application-specific) plan for the review of these 
applications for compliance with applicable HFE requirements. The ISG 
describes a process and provides implementation guidance for the NRC to 
tailor HFE review plans to each application to achieve an effective and 
efficient review.
    (3) The NRC has identified future guidance activities that would 
need to be completed after the part 53 proposed rule is published for 
public comment to support advanced reactor applications and NRC 
reviews. For example, the NRC recognizes that new guidance would be 
needed for the implementation of provisions in proposed Sec.  53.620(d) 
and the associated licensing provisions in proposed subpart H that 
would allow and establish requirements for the loading of fuel into a 
manufactured reactor for subsequent transport to and use at a 
commercial nuclear plant that will operate the facility pursuant to a 
COL. The NRC has not yet initiated the development of guidance 
documents in this category but will engage stakeholders during the 
development of these documents to ensure common prioritization. In 
addition, the NRC works with standards development organizations, 
advanced reactor developers, DOE, and other stakeholders to identify 
and facilitate new consensus codes and standards needed for advanced 
reactor development. The NRC will continue its membership and 
participation on standards development committees and working groups to 
support standards for advanced reactor technologies, where appropriate.

XVIII. Public Meeting

    The NRC will conduct a public meeting on this proposed rule for the 
purpose of describing the proposed rule and implementation guidance to 
the public and answering questions from the public on the proposed rule 
and implementation guidance.
    The NRC will publish a notice of the public meeting's location, 
time, and agenda on the NRC's public meeting website at least 10 
calendar days before the meeting. Stakeholders should monitor the NRC's 
public meeting website for information about the public meeting at: 
https://www.nrc.gov/public-involve/public-meetings/index.cfm.

XIX. Availability of Documents

    The documents identified in the following table are available to 
interested persons through one or more of the following methods, as 
indicated.

------------------------------------------------------------------------
                                                ADAMS accession No./Web
                   Document                      link/ Federal Register
                                                        Citation
------------------------------------------------------------------------
                         Proposed Rule Documents
------------------------------------------------------------------------
Federal Register Notification, ``Proposed      ML24095A161.
 Rule: Risk-Informed, Technology-Inclusive
 Regulatory Framework for Advanced
 Reactors,'' October, 2024.
``Draft Environmental Assessment for the       ML24095A163.
 Proposed Rule--Risk Informed, Technology-
 Inclusive Regulatory Framework for Advanced
 Reactors,'' October, 2024.
``Draft Regulatory Analysis for the Proposed   ML24095A166.
 Rule: Risk Informed, Technology-Inclusive
 Regulatory Framework for Advanced
 Reactors,'' October, 2024.
------------------------------------------------------------------------
                    Information Collection Documents
------------------------------------------------------------------------
Draft Supporting Statement for Information     ML21162A109.
 Collection Analysis--10 CFR Part 53.
Draft Supporting Statement for Information     ML23030A400.
 Collection Analysis--10 CFR Part 26.
Draft Supporting Statement for Information     ML24220A036.
 Collection Analysis--10 CFR Part 50.
Draft Supporting Statement for Information     ML23030A576.
 Collection Analysis--10 CFR Part 73.
Draft Supporting Statement for Information     ML24220A034.
 Collection Analysis--NRC Form 361S.
Draft Supporting Statement for Information     ML24220A035.
 Collection Analysis--NRC Form 366.
Draft Supporting Statement for Information     ML24220A033.
 Collection Analysis--NRC Form 893 and 894.
Proposed Rule--Part 26 Burden Tables for Risk  ML24240A008.
 Informed, Technology-Inclusive Regulatory
 Framework for Advanced Reactors.
Proposed Rule--Part 50 Burden Tables for Risk  ML24220A061.
 Informed, Technology-Inclusive Regulatory
 Framework for Advanced Reactors.
Proposed Rule--Part 53 Burden Tables for Risk  ML24220A060.
 Informed, Technology-Inclusive Regulatory
 Framework for Advanced Reactors.
Proposed Rule--Part 73 Burden Tables for Risk  ML24240A009.
 Informed, Technology-Inclusive Regulatory
 Framework for Advanced Reactors.
Draft NRC Form 361S, ``Part 53 Plant Event     ML23032A443.
 Notification Worksheet''.
Draft NRC Form 366, ``Licensee Event Report    ML23032A445.
 (LER)''.
Draft NRC Form 366A, ``Licensee Event Report   ML23032A447.
 (LER) Continuation Sheet''.
Draft NRC Form 366B, ``Licensee Event Report   ML23032A454.
 (LER) (Failure Continuation)''.
Draft NRC Form 893, ``10 CFR Part 26, Subpart  ML23032A435.
 M, Single FFD Policy Violation Form''.
Draft NRC Form 894, ``10 CFR Part 26, Subpart  ML23032A439.
 M, Annual Reporting Form for FFD Performance
 Information''.
------------------------------------------------------------------------
                   Draft Regulatory Guidance Documents
------------------------------------------------------------------------
DG-1413, ``Technology-Inclusive                ML22257A173.
 Identification Of Licensing Events For
 Commercial Nuclear Plants,'' October, 2024.
DG-5073, ``Fitness-For-Duty Programs For       ML22200A037.
 Commercial Nuclear Plants And Manufacturing
 Facilities Licensed Under 10 CFR Part 53,''
 October, 2024.
DG-5074, ``Access Authorization Program for    ML22199A246.
 Commercial Nuclear Plants,'' October, 2024.
DG-5075, ``Establishing Cybersecurity          ML22199A257.
 Programs For Commercial Nuclear Plants
 Licensed Under 10 CFR Part 53,'' October,
 2024.
DG-5076, ``Guidance for Technology Inclusive   ML22203A131.
 Requirements for Physical Protection of
 Licensed Activities at Commercial Nuclear
 Plants,'' October, 2024.

[[Page 87013]]

 
DG-5078, ``Fatigue Management For Nuclear      ML22264A109.
 Power Plant Personnel At Commercial Nuclear
 Plants Licensed Under 10 CFR Part 53,''
 October, 2024.
------------------------------------------------------------------------
                           Draft ISG Documents
------------------------------------------------------------------------
Draft ISG DRO-ISG-2023-01, ``Operator          ML22266A066.
 Licensing Programs,'' October, 2024.
Draft ISG DRO-ISG-2023-02, ``Interim Staff     ML22266A068.
 Guidance Augmenting NUREG-1791, `Guidance
 for Assessing Exemption Requests from the
 Nuclear Power Plant Licensed Operator
 Staffing Requirements Specified in 10 CFR
 50.54(m),' for Licensing Commercial Nuclear
 Plants under 10 CFR Part 53,'' October, 2024.
Draft ISG DRO-ISG-2023-03, ``Development of    ML22266A072.
 Scalable Human Factors Engineering Review
 Plans,'' October, 2024.
------------------------------------------------------------------------
                            Other References
------------------------------------------------------------------------
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[[Page 87014]]

 
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 2022.
SECY-83-293, ``Amendments to 10 CFR 50         ML21278A823 (non-public);
 Related to Anticipated Transients Without      ML21278A994 (non-
 Scram (ATWS) Events,'' dated July 19, 1983.    public).
SECY-93-092, ``Issues Pertaining to the        ML040210725.
 Advanced Reactor (PRISM, MHTGR, and PIUS)
 and CANDU 3 Designs and their Relationship
 to Current Regulatory Requirements,'' dated
 April 8, 1993.

[[Page 87015]]

 
SRM-SECY-10-0121, ``Modifying the              ML110610166.
 Risk[dash]Informed Regulatory Guidance for
 New Reactors,'' dated March 2, 2011.
SRM-SECY-17-0100, ``Security Baseline          ML18283A072.
 Inspection Program Assessment Results and
 Recommendations for Program Efficiencies,''
 dated October 8, 2018.
SRM-SECY-20-0032, ``Rulemaking Plan on         ML20276A293.
 `Risk[dash]Informed,
 Technology[dash]Inclusive Regulatory
 Framework for Advanced Reactors (RIN-3150-
 AK31; NRC-2019-0062),''' dated October 2,
 2020.
SRM-SECY-20-0045, ``Population Related Siting  ML22194A885.
 Considerations for Advanced Reactors,''
 dated July 30, 2022.
SRM-SECY-98-144, ``Staff Requirements--SECY-   ML003753593.
 98-144--White Paper on Risk-Informed and
 Performance-Based Regulations,'' dated
 February 24, 1999.
SECY-23-0021, ``Proposed Rule: Risk-Informed,  ML21162A095.
 Technology-Inclusive Regulatory Framework
 for Advanced Reactors (RIN 3150-AK31),''
 March 1, 2023.
SECY-23-0021, Enclosure 1, ``Draft Federal     ML21162A102.
 Register Notification''.
SECY-23-0021, Enclosure 2, ``Draft             ML21162A104.
 Environmental Assessment for the Proposed
 Rule--Risk Informed, Technology-Inclusive
 Regulatory Framework for Advanced Reactors''.
SECY-23-0021, Enclosure 3, ``Draft Regulatory  ML21165A112.
 Analysis for the Proposed Rule: Risk
 Informed, Technology-Inclusive Regulatory
 Framework for Advanced Reactors''.
SECY-23-0021, Enclosure 4, ``Alternative       ML22244A001.
 Approaches Considered for Selected Topics
 During the Development of 10 CFR Part 53''.
SECY-23-0021, Enclosure 5, ``Estimated         ML22304A099 (non-public).
 Resources for The Risk-Informed, Technology-
 Inclusive Regulatory Framework For Advanced
 Reactors Rulemaking''.
Staff Requirements--SECY-23-0021, ``Proposed   ML24064A047 (package).
 Rule: Risk-Informed, Technology-Inclusive
 Regulatory Framework for Advanced Reactors
 (RIN 3150-AK31),'' March 4, 2024.
------------------------------------------------------------------------

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

List of Subjects

10 CFR Part 1

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

10 CFR Part 2

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

10 CFR Part 10

    Administrative practice and procedure, Classified information, 
Government employees, Security measures.

10 CFR Part 11

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

10 CFR Part 19

    Criminal penalties, Environmental protection, Nuclear Energy, 
Nuclear materials, Nuclear power plants and reactors, Occupational 
safety and health, Penalties, Radiation protection, Reporting and 
recordkeeping requirements, Sex discrimination.

10 CFR Part 20

    Byproduct material, Criminal penalties, Hazardous waste, Licensed 
material, Nuclear energy, Nuclear materials, Nuclear power plants and 
reactors, Occupational safety and health, Packaging and containers, 
Penalties, Radiation protection, Reporting and recordkeeping 
requirements, Source material, Special nuclear material, Waste 
treatment and disposal.

10 CFR Part 21

    Nuclear power plants and reactors, Penalties, Radiation protection, 
Reporting and recordkeeping requirements.

10 CFR Part 25

    Classified information, Criminal penalties, Investigations, 
Penalties, Reporting and recordkeeping requirements, Security measures.

10 CFR Part 26

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

10 CFR Part 30

    Byproduct material, Criminal penalties, Government contracts, 
Intergovernmental relations, Isotopes, Nuclear energy, Nuclear 
materials, Penalties, Radiation protection, Reporting and recordkeeping 
requirements, Whistleblowing.

10 CFR Part 40

    Criminal penalties, Exports, Government contracts, Hazardous 
materials transportation, Hazardous waste, Nuclear energy, Nuclear 
materials, Penalties, Reporting and recordkeeping requirements, Source 
material, Uranium, Whistleblowing.

10 CFR Part 50

    Administrative practice and procedure, Antitrust, Backfitting, 
Classified information, Criminal penalties, Education, Emergency 
planning, Fire prevention, Fire protection, Intergovernmental 
relations, Nuclear power plants and reactors, Penalties, Radiation 
protection, Reactor siting criteria, Reporting and

[[Page 87016]]

recordkeeping requirements, Whistleblowing.

10 CFR Part 51

    Administrative practice and procedure, Environmental impact 
statements, Hazardous waste, Nuclear energy, Nuclear materials, Nuclear 
power plants and reactors, Reporting and recordkeeping requirements.

10 CFR Part 53

    Administrative practice and procedure, Antitrust, Backfitting, 
Construction permit, Combined license, Classified information, Criminal 
penalties, Early site permit, Emergency planning, Fees, Fire 
prevention, Fire protection, Inspection, Intergovernmental relations, 
Limited work authorization, Manufacturing license, Nuclear power plants 
and reactors, Operating license, Penalties, Prototype, Radiation 
protection, Reactor siting criteria, Reporting and recordkeeping 
requirements, Standard design, Standard design certification, Training 
programs.

10 CFR Part 70

    Classified information, Criminal penalties, Emergency medical 
services, Hazardous materials transportation, Material control and 
accounting, Nuclear energy, Nuclear materials, Packaging and 
containers, Penalties, Radiation protection, Reporting and 
recordkeeping requirements, Scientific equipment, Security measures, 
Special nuclear material, Whistleblowing.

10 CFR Part 72

    Administrative practice and procedure, Hazardous waste, Indians, 
Intergovernmental relations, Nuclear energy, Penalties, Radiation 
protection, Reporting and recordkeeping requirements, Security 
measures, Spent fuel, Whistleblowing.

10 CFR Part 73

    Criminal penalties, Exports, Hazardous materials transportation, 
Imports, Nuclear energy, Nuclear materials, Nuclear power plants and 
reactors, Penalties, Reporting and recordkeeping requirements, Security 
measures.

10 CFR Part 74

    Accounting, Criminal penalties, Hazardous materials transportation, 
Material control and accounting, Nuclear energy, Nuclear materials, 
Packaging and containers, Penalties, Radiation protection, Reporting 
and recordkeeping requirements, Scientific equipment, Special nuclear 
material.

10 CFR Part 75

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

10 CFR Part 95

    Classified information, Criminal penalties, Penalties, Reporting 
and recordkeeping requirements, Security measures.

10 CFR Part 140

    Criminal penalties, Extraordinary nuclear occurrence, Insurance, 
Intergovernmental relations, Nuclear materials, Nuclear power plants 
and reactors, Penalties, Reporting and recordkeeping requirements.

10 CFR Part 150

    Criminal penalties, Hazardous materials transportation, 
Intergovernmental relations, Nuclear energy, Nuclear materials, 
Penalties, Reporting and recordkeeping requirements, Security measures, 
Source material, Special nuclear material.

10 CFR Part 170

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

10 CFR Part 171

    Annual charges, Approvals, Byproduct material, Holders of 
certificates, Intergovernmental relations, Nonpayment penalties, 
Nuclear materials, Nuclear power plants and reactors, Registrations, 
Source material, Special nuclear material.

    For the reasons set out in the preamble and under the authority of 
the Atomic Energy Act of 1954, as amended; the Energy Reorganization 
Act of 1974, as amended; and 5 U.S.C. 552 and 553, the NRC is proposing 
the following amendments to 10 CFR parts 1, 2, 10, 11, 19, 20, 21, 25, 
26, 30, 40, 50, 51, 70, 72, 73, 74, 75, 95, 140, 150, 170, and 171 and 
adding 10 CFR part 53:

PART 1--STATEMENT OF ORGANIZATION AND GENERAL INFORMATION

0
1. The authority citation for part 1 continues to read as follows:

    Authority:  Atomic Energy Act of 1954, secs. 23, 25, 29, 161, 
191 (42 U.S.C. 2033, 2035, 2039, 2201, 2241); Energy Reorganization 
Act of 1974, secs. 201, 203, 204, 205, 209 (42 U.S.C. 5841, 5843, 
5844, 5845, 5849); Administrative Procedure Act (5 U.S.C. 552, 553); 
Reorganization Plan No. 1 of 1980, 5 U.S.C. Appendix (Reorganization 
Plans).


Sec.  1.43  [Amended]

0
2. In Sec.  1.43, in paragraph (a)(2) remove the cross reference ``10 
CFR parts 50, 52, and 54'' and add in its place the cross reference 
``10 CFR parts 50, 52, 53, and 54''.

PART 2--AGENCY RULES OF PRACTICE AND PROCEDURE

0
3. The authority citation for part 2 continues to read as follows:

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


Sec.  2.1  [Amended]

0
4. In Sec.  2.1, in paragraph (e) remove the phrase ``part 52'' and add 
in its place the phrase ``part 52 or part 53''.
0
5. In Sec.  2.4, revise the definitions for ``Contested proceeding'' 
and ``Facility'' to read as follows:


Sec.  2.4  Definitions.

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

[[Page 87017]]

Sec.  2.100  [Amended]

0
6. In Sec.  2.100, remove the phrase ``subpart E of part 52'' and add 
in its place the phrase ``subpart E of part 52 or subpart H of part 
53''.
0
7. In Sec.  2.101, revise paragraphs (a)(3)(i), (a)(5), (a)(9) 
introductory text and paragraph (a)(9)(i) to read as follows:


Sec.  2.101  Filing of application.

    (a) * * *
    (3) * * *
    (i) Submit to the Director, Office of Nuclear Reactor Regulation, 
or Director, Office of Nuclear Material Safety and Safeguards, as 
appropriate, such additional copies as the regulations in part 50, 
subpart A of part 51, and part 53 of this chapter require;
* * * * *
    (5) An applicant for a construction permit under parts 50 or 53 of 
this chapter or a combined license under parts 52 or 53 of this chapter 
for a production or utilization facility which is subject to Sec.  
51.20(b) of this chapter, and is of the type specified in Sec.  
50.21(b)(2) or (b)(3); or Sec.  50.22; or part 53, as applicable, of 
this chapter, or is a testing facility, may submit the information 
required of applicants by parts 50, 52, or 53 of this chapter in two 
parts. One part shall be accompanied by the information required by 
Sec.  50.30(f) of this chapter, Sec.  52.80(b) of this chapter, or 
Sec.  53.1100(f) of this chapter, as applicable. The other part shall 
include any information required by Sec.  50.34(a) and, if applicable, 
Sec.  50.34a of this chapter; or Sec. Sec.  52.79 and 52.80(a) of this 
chapter; or Sec. Sec.  53.1109, 53.1306, 53.1309, and 53.1312 of this 
chapter; or Sec. Sec.  53.1109, 53.1413, 53.1416, and 53.1419 of this 
chapter, as applicable. One part may precede or follow other parts by 
no longer than 6 months. If it is determined that either of the parts 
as described above is incomplete and not acceptable for processing, the 
Director, Office of Nuclear Reactor Regulation, or Director, Office of 
Nuclear Material Safety and Safeguards, as appropriate, will inform the 
applicant of this determination and the respects in which the document 
is deficient. Such a determination of completeness will generally be 
made within a period of 30 days. Whichever part is filed first shall 
also include the fee required by Sec.  50.30(e) or Sec.  53.1100(e) and 
Sec.  170.21 of this chapter and the information required by Sec. Sec.  
50.33, 50.34(a)(1), and 52.79(a)(1) of this chapter; or Sec. Sec.  
53.1109, 53.1309, and 53.1416 of this chapter, as applicable, and Sec.  
50.37 or Sec.  53.1115, as applicable, of this chapter. The Director, 
Office of Nuclear Reactor Regulation, or Director, Office of Nuclear 
Material Safety and Safeguards, as appropriate, will accept for 
docketing an application for a construction permit under part 50 or 
part 53 of this chapter or a combined license under parts 52 or 53 of 
this chapter for a production or utilization facility which is subject 
to Sec.  51.20(b) of this chapter, and is of the type specified in 
Sec.  50.21(b)(2) or (b)(3), or Sec.  50.22, or part 53, as applicable, 
of this chapter or is a testing facility where one part of the 
application as described above is complete and conforms to the 
requirements of part 50 of this chapter. The additional parts will be 
docketed upon a determination by the Director, Office of Nuclear 
Reactor Regulation, or Director, Office of Nuclear Material Safety and 
Safeguards, as appropriate, that it is complete.
* * * * *
    (9) An applicant for a construction permit for a utilization 
facility which is subject to Sec.  51.20(b) of this chapter and is of 
the type specified in Sec.  50.21(b)(2) or (b)(3), or Sec.  50.22, or 
part 53 of this chapter, an applicant for or holder of an early site 
permit under part 52 or part 53 of this chapter, or an applicant for a 
combined license under parts 52 or 53 of this chapter, who seeks to 
conduct the activities authorized under Sec.  50.10(d) or Sec.  53.1130 
of this chapter may submit a complete application under paragraphs 
(a)(1) through (a)(4) of this section which includes the information 
required by Sec.  50.10(d) or Sec.  53.1130 of this chapter. 
Alternatively, the applicant (other than an applicant for or holder of 
an early site permit) may submit its application in two parts:
    (i) Part one must include the information required by Sec.  
50.33(a) through (f) or Sec.  53.1109(a) through (e) and Sec.  53.1306 
of this chapter, and the information required by Sec.  50.10(d)(2) and 
(d)(3) or Sec.  53.1130(a)(2) and (a)(3) of this chapter, as 
applicable.
* * * * *
0
8. In Sec.  2.104, revise paragraph (a) to read as follows:


Sec.  2.104  Notice of hearing.

    (a) In the case of an application on which a hearing is required by 
the Act or this chapter, or in which the Commission finds that a 
hearing is required in the public interest, the Secretary will issue a 
notice of hearing to be published in the Federal Register. The notice 
must be published at least 15 days, and in the case of an application 
concerning a limited work authorization, construction permit, early 
site permit, or combined license for a facility of the type described 
in Sec.  50.21(b) or 50.22, or subpart H of part 53 of this chapter, as 
applicable, or a testing facility, at least 30 days, before the date 
set for hearing in the notice.\1\ In addition, in the case of an 
application for a limited work authorization, construction permit, 
early site permit, or combined license for a facility of the type 
described in Sec.  50.22 or subpart H of part 53 of this chapter, as 
applicable, or a testing facility, the notice must be issued as soon as 
practicable after the NRC has docketed the application. If the 
Commission decides, under Sec.  2.101(a)(2), to determine the 
acceptability of the application based on its technical adequacy as 
well as completeness, the notice must be issued as soon as practicable 
after the application has been tendered.
* * * * *
    \1\ If the notice of hearing concerning an application for a 
limited work authorization, construction permit, early site permit, 
or combined license for a facility of the type described in Sec.  
50.21(b) or Sec.  50.22, or subpart H of part 53 of this chapter, as 
applicable, or a testing facility, does not specify the time and 
place of initial hearing, a subsequent notice will be published in 
the Federal Register which will provide at least 30-day notice of 
the time and place of that hearing. After this notice is given, the 
presiding officer may reschedule the commencement of the initial 
hearing for a later date or reconvene a recessed hearing without 
again providing at least 30-day notice.

0
9. In Sec.  2.105, revise paragraph (a) introductory text and 
paragraphs (a)(4), (a)(10), (a)(12), (a)(13), (b)(3) introductory text, 
(b)(3)(i), (ii), and (iv) to read as follows:


Sec.  2.105  Notice of proposed action.

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

[[Page 87018]]

chapter that the amendment involves no significant hazards 
consideration, though it will provide notice of opportunity for a 
hearing pursuant to this section, it may make the amendment immediately 
effective and grant a hearing thereafter; or
    (ii) If the Commission determines under Sec. Sec.  50.58 and 50.91 
or Sec.  53.1515 of this chapter, as applicable, that an emergency 
situation exists or that exigent circumstances exist and that the 
amendment involves no significant hazards consideration, it will 
provide notice of opportunity for a hearing pursuant to Sec.  2.106 (if 
a hearing is requested, it will be held after issuance of the 
amendment);
* * * * *
    (10) In the case of an application for an operating license for a 
facility of a type described in Sec.  50.21(b) or Sec.  50.22, or part 
53 of this chapter or a testing facility, a notice of opportunity for 
hearing shall be issued as soon as practicable after the application 
has been docketed; or
* * * * *
    (12) An amendment to an early site permit issued under subpart A of 
part 52, or under subpart H of part 53 of this chapter, as follows:
    (i) If the early site permit does not provide authority to conduct 
the activities allowed under Sec.  50.10(e)(1) or Sec.  53.1130(b)(1) 
of this chapter, the amendment will involve no significant hazards 
consideration, and though the NRC will provide notice of opportunity 
for a hearing under this section, it may make the amendment immediately 
effective and grant a hearing thereafter; and
    (ii) If the early site permit provides authority to conduct the 
activities allowed under Sec.  50.10(e)(1) or Sec.  53.1130(b)(1) of 
this chapter and the Commission determines under Sec. Sec.  50.58 and 
50.91 or Sec.  53.1515 of this chapter that an emergency situation 
exists or that exigent circumstances exist and that the amendment 
involves no significant hazards consideration, it will provide notice 
of opportunity for a hearing under Sec.  2.106 of this chapter (if a 
hearing is requested, which will be held after issuance of the 
amendment).
    (13) A manufacturing license under subpart F of part 52 or subpart 
H of part 53 of this chapter.
    (b) * * *
    (3) For a notice of intended operation under Sec.  52.103(a) or 
Sec.  53.1452(a) of this chapter, the following information:
    (i) The identification of the NRC action as making the finding 
required under Sec.  52.103(g) or Sec.  53.1452(g) of this chapter;
    (ii) The manner in which the licensee notifications under Sec.  
52.99(c) or Sec.  53.1449(c), of this chapter which are required to be 
made available by Sec.  52.99(e)(2) or Sec.  53.1449(e)(2), of this 
chapter may be obtained and examined;
* * * * *
    (iv) Any conditions, limitations, or restrictions to be placed on 
the license in connection with the finding under Sec.  52.103(g) or 
Sec.  53.1452(g) of this chapter, and the expiration date or 
circumstances (if any) under which the conditions, limitations or 
restrictions will no longer apply.
* * * * *
0
10. In Sec.  2.106, revise paragraphs (a)(2), (a)(3), and (b)(2) 
introductory text to read as follows:


Sec.  2.106  Notice of issuance.

    (a) * * *
    (2) An amendment of a license for a facility of the type described 
in Sec.  50.21(b) or Sec.  50.22, or part 53 of this chapter, as 
applicable, or a testing facility, whether or not a notice of proposed 
action has been previously published; and
    (3) The finding under Sec.  52.103(g) or Sec.  53.1452(g) of this 
chapter.
    (b) * * *
    (2) In the case of a finding under Sec.  52.103(g) or Sec.  
53.1452(g) of this chapter:
* * * * *
0
11. In Sec.  2.109, revise paragraphs (b), (c), and (d) to read as 
follows:


Sec.  2.109  Effect of timely renewal application.

* * * * *
    (b) If the licensee of a nuclear power plant licensed under Sec.  
50.21(b) or Sec.  50.22 or under subpart H of part 53 of this chapter 
files a sufficient application for renewal of either an operating 
license or a combined license at least 5 years before the expiration of 
the existing license, the existing license will not be deemed to have 
expired until the application has been finally determined.
    (c) If the holder of an early site permit licensed under subpart A 
of part 52 or under subpart H of part 53 of this chapter, as 
applicable, files a sufficient application for renewal under Sec.  
52.29 or Sec.  53.1173 of this chapter, as applicable, at least 12 
months before the expiration of the existing early site permit, the 
existing permit will not be deemed to have expired until the 
application has been finally determined.
    (d) If the licensee of a manufacturing license under subpart F of 
part 52, or under subpart H of part 53 of this chapter files a 
sufficient application for renewal under Sec.  52.177 or Sec.  53.1295 
of this chapter at least 12 months before the expiration of the 
existing license, the existing license will not be deemed to have 
expired until the application has been finally determined.
* * * * *
0
12. In Sec.  2.110, revise paragraphs (a)(1) and (b) to read as 
follows:


Sec.  2.110  Filing and administrative action on submittals for 
standard design approval or early review of site suitability issues.

    (a)(1) A submittal for a standard design approval under subpart E 
of part 52 or under subpart H of part 53 of this chapter shall be 
subject to Sec. Sec.  2.101(a) and 2.390 to the same extent as if it 
were an application for a permit or license.
* * * * *
    (b) Upon initiation of review by the NRC staff of a submittal for 
an early review of site suitability issues under appendix Q to part 50 
of this chapter, or for a standard design approval under subpart E of 
part 52 or under subpart H of part 53 of this chapter, the Director, 
Office of Nuclear Reactor Regulation, shall publish in the Federal 
Register a notice of receipt of the submittal, inviting comments from 
interested persons within 60 days of publication or other time as may 
be specified, for consideration by the NRC staff and ACRS in their 
review.
* * * * *
0
13. In Sec.  2.202, revise paragraph (e) to read as follows:


Sec.  2.202  Orders.

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

[[Page 87019]]

the requirements of Sec.  52.63 or Sec.  53.1263 of this chapter, as 
applicable, must be followed, unless the applicant or licensee has 
consented to follow the action required.
    (5) If the order involves a change to a standard design approval 
referenced by that plant's application, the requirements of Sec.  
52.145 or Sec.  53.1221 of this chapter, as applicable, must be 
followed unless the applicant or licensee has consented to follow the 
action required.
    (6) If the order involves a modification of a manufacturing license 
under subpart F of part 52 or under subpart H of part 53 of this 
chapter, the requirements of Sec.  52.171 or Sec.  53.1288 of this 
chapter, as applicable, must be followed, unless the applicant or 
licensee has consented to the action required.
0
14. In Sec.  2.309, revise paragraphs (a), (f)(1)(i), (f)(1)(vi) and 
(vii), (g), (h)(2), (i)(2), (j) to read as follows:


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

    (a) General requirements. Any person whose interest may be affected 
by a proceeding and who desires to participate as a party must file a 
written request for hearing and a specification of the contentions 
which the person seeks to have litigated in the hearing. In a 
proceeding under Sec.  52.103 or Sec.  53.1452 of this chapter, as 
applicable, the Commission, acting as the presiding officer, will grant 
the request if it determines that the requestor has standing under the 
provisions of paragraph (d) of this section and has proposed at least 
one admissible contention that meets the requirements of paragraph (f) 
of this section. For all other proceedings, except as provided in 
paragraph (e) of this section, the Commission, presiding officer, or 
the Atomic Safety and Licensing Board designated to rule on the request 
for hearing and/or petition for leave to intervene, will grant the 
request/petition if it determines that the requestor/petitioner has 
standing under the provisions of paragraph (d) of this section and has 
proposed at least one admissible contention that meets the requirements 
of paragraph (f) of this section. In ruling on the request for hearing/
petition to intervene submitted by petitioners seeking to intervene in 
the proceeding on the HLW repository, the Commission, the presiding 
officer, or the Atomic Safety and Licensing Board shall also consider 
any failure of the petitioner to participate as a potential party in 
the pre-license application phase under subpart J of this part in 
addition to the factors in paragraph (d) of this section. If a request 
for hearing or petition to intervene is filed in response to any notice 
of hearing or opportunity for hearing, the applicant/licensee shall be 
deemed to be a party.
* * * * *
    (f) * * *
    (1) * * *
    (i) Provide a specific statement of the issue of law or fact to be 
raised or controverted, provided further, that the issue of law or fact 
to be raised in a request for hearing under Sec.  52.103(b) or Sec.  
53.1452(b) of this chapter, as applicable, must be directed at 
demonstrating that one or more of the acceptance criteria in the 
combined license have not been, or will not be met, and that the 
specific operational consequences of nonconformance would be contrary 
to providing reasonable assurance of adequate protection of the public 
health and safety;
* * * * *
    (vi) In a proceeding other than one under Sec.  52.103 or Sec.  
53.1452 of this chapter provide sufficient information to show that a 
genuine dispute exists with the applicant/licensee on a material issue 
of law or fact. This information must include references to specific 
portions of the application (including the applicant's environmental 
report and safety report) that the petitioner disputes and the 
supporting reasons for each dispute, or, if the petitioner believes 
that the application fails to contain information on a relevant matter 
as required by law, the identification of each failure and the 
supporting reasons for the petitioner's belief; and
    (vii) In a proceeding under Sec.  52.103(b) or Sec.  53.1452(b) of 
this chapter, as applicable, the information must be sufficient, and 
include supporting information showing, prima facie, that one or more 
of the acceptance criteria in the combined license have not been, or 
will not be met, and that the specific operational consequences of 
nonconformance would be contrary to providing reasonable assurance of 
adequate protection of the public health and safety. This information 
must include the specific portion of the report required by Sec.  
52.99(c) or Sec.  53.1449(c) of this chapter, as applicable, which the 
requestor believes is inaccurate, incorrect, and/or incomplete (i.e., 
fails to contain the necessary information required by Sec.  52.99(c) 
or Sec.  53.1449(c) of this chapter, as applicable). If the requestor 
identifies a specific portion of the report under Sec.  52.99(c) or 
Sec.  53.1449(c) of this chapter, as applicable, as incomplete and the 
requestor contends that the incomplete portion prevents the requestor 
from making the necessary prima facie showing, then the requestor must 
explain why this deficiency prevents the requestor from making the 
prima facie showing.
* * * * *
    (g) Selection of hearing procedures. A request for hearing and/or 
petition for leave to intervene may, except in a proceeding under Sec.  
52.103 or Sec.  53.1452 of this chapter, as applicable, also address 
the selection of hearing procedures, taking into account the provisions 
of Sec.  2.310. If a request/petition relies upon Sec.  2.310(d), the 
request/petition must demonstrate, by reference to the contention and 
the bases provided and the specific procedures in subpart G of this 
part, that resolution of the contention necessitates resolution of 
material issues of fact which may be best determined through the use of 
the identified procedures.
    (h) * * *
    (2) If the proceeding pertains to a production or utilization 
facility (as defined in Sec.  50.2 or Sec.  53.020 of this chapter) 
located within the boundaries of the State, local governmental body, or 
Federally-recognized Indian Tribe seeking to participate as a party, no 
further demonstration of standing is required. If the production or 
utilization facility is not located within the boundaries of the State, 
local governmental body, or Federally-recognized Indian Tribe seeking 
to participate as a party, the State, local governmental body, or 
Federally-recognized Indian Tribe also must demonstrate standing.
* * * * *
    (i) * * *
    (2) Except in a proceeding under Sec.  52.103 or Sec.  53.1452 of 
this chapter, as applicable, the participant who filed the hearing 
request, intervention petition, or motion for leave to file new or 
amended contentions after the deadline may file a reply to any answer. 
The reply must be filed within 7 days after service of that answer.
* * * * *
    (j) Decision on request/petition. (1) In all proceedings other than 
a proceeding under Sec.  52.103 or Sec.  53.1452 of this chapter, as 
applicable, the presiding officer shall issue a decision on each 
request for hearing or petition to intervene within 45 days of the 
conclusion of the initial pre-hearing conference or, if no pre-hearing 
conference is conducted, within 45 days after the filing of answers and 
replies

[[Page 87020]]

under paragraph (i) of this section. With respect to a request to admit 
amended or new contentions, the presiding officer shall issue a 
decision on each such request within 45 days of the conclusion of any 
pre-hearing conference that may be conducted regarding the proposed 
amended or new contentions or, if no pre-hearing conference is 
conducted, within 45 days after the filing of answers and replies, if 
any. In the event the presiding officer cannot issue a decision within 
45 days, the presiding officer shall issue a notice advising the 
Commission and the parties, and the notice shall include the expected 
date of when the decision will issue.
    (2) The Commission, acting as the presiding officer, shall 
expeditiously grant or deny the request for hearing in a proceeding 
under Sec.  52.103 or Sec.  53.1452 of this chapter, as applicable. The 
Commission's decision may not be the subject of any appeal under Sec.  
2.311.
0
15. Amend Sec.  2.310 by:
0
a. In paragraphs (a) and (h) introductory text, removing the cross-
reference ``parts 30, 32 through 36, 39, 40, 50, 52, 54, 55, 61, 70 and 
72 of this chapter'' and adding, in its place, the cross reference 
``parts 30, 32 through 36, 39, 40, 50, 52, 53, 54, 55, 61, 70 and 72 of 
this chapter''; and
0
b. Revising paragraphs (i) and (j).
    The revisions read as follows.


Sec.  2.310  Selection of hearing procedures.

* * * * *
    (i) In design certification rulemaking proceedings under part 52 or 
part 53 of this chapter, any informal hearing held under Sec.  52.51 or 
Sec.  53.1242 of this chapter, as applicable, must be conducted under 
the procedures of subpart O of this part.
    (j) Proceedings on a Commission finding under Sec.  52.103(c) and 
(g) or Sec.  53.1452(c) and (g) of this chapter, as applicable, shall 
be conducted in accordance with the procedures designated by the 
Commission in each proceeding.
* * * * *
0
16. In Sec.  2.329, revise paragraph (a) to read as follows:


Sec.  2.329  Prehearing conference.

    (a) Necessity for prehearing conference; timing. The Commission or 
the presiding officer may, and in the case of a proceeding on an 
application for a construction permit or an operating license for a 
facility of a type described in Sec. Sec.  50.21(b) or 50.22, or part 
53 of this chapter, or a testing facility, must direct the parties or 
their counsel to appear at a specified time and place for a conference 
or conferences before trial. A prehearing conference in a proceeding 
involving a construction permit or operating license for a facility of 
a type described in Sec. Sec.  50.21(b) or 50.22 or part 53 of this 
chapter must be held within sixty (60) days after discovery has been 
completed or any other time specified by the Commission or the 
presiding officer.
* * * * *
0
17. In Sec.  2.339, revise paragraph (d) to read as follows:


Sec.  2.339  Expedited decision-making procedure.

* * * * *
    (d) The provisions of this section do not apply to an initial 
decision directing the issuance of a limited work authorization under 
10 CFR 50.10 or 10 CFR 53.1130; an early site permit under subpart A of 
part 52 or under subpart H of part 53 of this chapter; a construction 
permit or construction authorization under part 50 or part 53 of this 
chapter; a combined license under subpart C of part 52 or under subpart 
H of part 53 of this chapter; or a manufacturing license under subpart 
F of part 52 or under subpart H of part 53.
0
18. In Sec.  2.340, revise paragraphs (b), (c), (d), (f), (i), and (j) 
to read as follows:


Sec.  2.340  Initial decision in certain contested proceedings; 
immediate effectiveness of initial decisions; issuance of 
authorizations, permits and licenses.

* * * * *
    (b) Initial decision--combined license under 10 CFR parts 52 or 53. 
(1) Matters in controversy; presiding officer consideration of matters 
not put in controversy by parties. In any initial decision in a 
contested proceeding on an application for a combined license under 
parts 52 or 53 of this chapter (including an amendment to or renewal of 
combined license), the presiding officer shall make findings of fact 
and conclusions of law on the matters put into controversy by the 
parties and any matter designated by the Commission to be decided by 
the presiding officer. The presiding officer shall also make findings 
of fact and conclusions of law on any matter not put into controversy 
by the parties, but only to the extent that the presiding officer 
determines that a serious safety, environmental, or common defense and 
security matter exists, and the Commission approves of an examination 
of and decision on the matter upon its referral by the presiding 
officer under, inter alia, the provisions of Sec. Sec.  2.323 and 
2.341.
    (2) Presiding officer initial decision and issuance of permit or 
license.
    (i) In a contested proceeding for the initial issuance or renewal 
of a combined license under parts 52 or 53 of this chapter, or the 
amendment of a combined license where the NRC has not made a 
determination of no significant hazards consideration, the Commission 
or the Director, Office of Nuclear Reactor Regulation, as appropriate 
after making the requisite findings, shall issue, deny, or 
appropriately condition the permit or license in accordance with the 
presiding officer's initial decision once that decision becomes 
effective.
    (ii) In a contested proceeding for the amendment of a combined 
license under parts 52 or 53 of this chapter where the NRC has made a 
determination of no significant hazards consideration, the Commission 
or the Director, Office of Nuclear Reactor Regulation, as appropriate 
(appropriate official), after making the requisite findings and 
complying with any applicable provisions of Sec.  2.1202(a) or Sec.  
2.1403(a), may issue the amendment before the presiding officer's 
initial decision becomes effective. Once the presiding officer's 
initial decision becomes effective, the appropriate official shall take 
action with respect to that amendment in accordance with the initial 
decision. If the presiding officer's initial decision becomes effective 
before the appropriate official issues the amendment, then the 
appropriate official, after making the requisite findings, shall issue, 
deny, or appropriately condition the amendment in accordance with the 
presiding officer's initial decision.
    (c) Initial decision on findings under 10 CFR 52.103 or 10 CFR 
53.1452 with respect to acceptance criteria in nuclear power reactor 
combined licenses. In any initial decision under Sec.  52.103(g) or 
Sec.  53.1452(g) of this chapter with respect to whether acceptance 
criteria have been or will be met, the presiding officer shall make 
findings of fact and conclusions of law on the matters put into 
controversy by the parties, and any matter designated by the Commission 
to be decided by the presiding officer. Matters not put into 
controversy by the parties but identified by the presiding officer as 
matters requiring further examination, shall be referred to the 
Commission for its determination; the Commission may, in its 
discretion, treat any of these referred matters as a request for action 
under Sec.  2.206 and process the matter in accordance with Sec.  
52.103(f) or Sec.  53.1452(f) of this chapter.
    (d) Initial decision--manufacturing license under 10 CFR parts 52 
or 53. (1) Matters in controversy; presiding officer consideration of 
matters not put in controversy by parties. In any initial decision in a 
contested proceeding on an application for a manufacturing

[[Page 87021]]

license under subpart C of part 52 or subpart H of part 53 of this 
chapter (including an amendment to or renewal of a manufacturing 
license), the presiding officer shall make findings of fact and 
conclusions of law on the matters put into controversy by the parties 
and any matter designated by the Commission to be decided by the 
presiding officer. The presiding officer also shall make findings of 
fact and conclusions of law on any matter not put into controversy by 
the parties, but only to the extent that the presiding officer 
determines that a serious safety, environmental, or common defense and 
security matter exists, and the Commission approves of an examination 
of and decision on the matter upon its referral by the presiding 
officer under, inter alia, the provisions of Sec. Sec.  2.323 and 
2.341.
    (2) Presiding officer initial decision and issuance of permit or 
license.
    (i) In a contested proceeding for the initial issuance or renewal 
of a manufacturing license under subpart C of part 52 or subpart H of 
part 53 of this chapter, or the amendment of a manufacturing license, 
the Commission or the Director, Office of Nuclear Reactor Regulation, 
as appropriate, after making the requisite findings, shall issue, deny, 
or appropriately condition the permit or license in accordance with the 
presiding officer's initial decision once that decision becomes 
effective.
    (ii) In a contested proceeding for the initial issuance or renewal 
of a manufacturing license under subpart C of part 52 or subpart H of 
part 53 of this chapter, or the amendment of a manufacturing license, 
the Commission or the Director, Office of Nuclear Reactor Regulation, 
as appropriate (appropriate official), may issue the license, permit, 
or license amendment in accordance with Sec.  2.1202(a) or Sec.  
2.1403(a) before the presiding officer's initial decision becomes 
effective. If, however, the presiding officer's initial decision 
becomes effective before the license, permit, or license amendment is 
issued under Sec.  2.1202 or Sec.  2.1403, then the Commission or the 
Director, Office of Nuclear Reactor Regulation, as appropriate, shall 
issue, deny, or appropriately condition the license, permit, or license 
amendment in accordance with the presiding officer's initial decision.
* * * * *
    (f) Immediate effectiveness of certain presiding officer decisions. 
A presiding officer's initial decision directing the issuance or 
amendment of a limited work authorization under Sec.  50.10 or Sec.  
53.1130 of this chapter; an early site permit under subpart A of part 
52 or under subpart H of part 53 of this chapter; a construction permit 
or construction authorization under part 50 or part 53 of this chapter; 
an operating license under part 50 or part 53 of this chapter; a 
combined license under subpart C of part 52 or subpart H or part 53 of 
this chapter; a manufacturing license under subpart F of part 52 or 
subpart H of part 53 of this chapter; a renewed license under part 53 
or part 54 of this chapter; or a license under part 72 of this chapter 
to store spent fuel in an independent spent fuel storage facility 
(ISFSI) or a monitored retrievable storage installation (MRS); an 
initial decision directing issuance of a license under part 61 of this 
chapter; or an initial decision under Sec.  52.103(g) or Sec.  
53.1452(g) of this chapter that acceptance criteria in a combined 
license have been met, is immediately effective upon issuance unless 
the presiding officer finds that good cause has been shown by a party 
why the initial decision should not become immediately effective.
* * * * *
    (i) Issuance of authorizations, permits, and licenses--production 
and utilization facilities. The Commission or the Director, Office of 
Nuclear Reactor Regulation, as appropriate, shall issue a limited work 
authorization under Sec.  50.10 or Sec.  53.1130 of this chapter; an 
early site permit under subpart A of part 52 or subpart H of part 53 of 
this chapter; a construction permit or construction authorization under 
part 50 or part 53 of this chapter; an operating license under part 50 
or part 53 of this chapter; a combined license under subpart C of part 
52 or part 53 of this chapter; or a manufacturing license under subpart 
F of part 52 or part 53 of this chapter within 10 days from the date of 
issuance of the initial decision:
    (1) If the Commission or the Director has made all findings 
necessary for issuance of the authorization, permit or license, not 
within the scope of the initial decision of the presiding officer; and
    (2) Notwithstanding the pendency of a petition for reconsideration 
under Sec.  2.345, a petition for review under Sec.  2.341, or a motion 
for stay under Sec.  2.342, or the filing of a petition under Sec.  
2.206.
    (j) Issuance of finding on acceptance criteria under 10 CFR 52.103 
or 10 CFR 53.1452. The Commission or the Director, Office of Nuclear 
Reactor Regulation, as appropriate, shall make the finding under Sec.  
52.103(g) or Sec.  53.1452(g) of this chapter, that acceptance criteria 
in a combined license are met within 10 days from the date of the 
presiding officer's initial decision:
    (1) If the Commission or the Director is otherwise able to make the 
finding under Sec.  52.103(g) or Sec.  53.1452(g) of this chapter, that 
the prescribed acceptance criteria are met for those acceptance 
criteria not within the scope of the initial decision of the presiding 
officer;
    (2) If the presiding officer's initial decision--with respect to 
contentions that the prescribed acceptance criteria have not been met--
finds that those acceptance criteria have been met, and the Commission 
or the Director thereafter is able to make the finding that those 
acceptance criteria are met;
    (3) If the presiding officer's initial decision--with respect to 
contentions that the prescribed acceptance criteria will not be met--
finds that those acceptance criteria will be met, and the Commission or 
the Director thereafter is able to make the finding that those 
acceptance criteria are met; and
    (4) Notwithstanding the pendency of a petition for reconsideration 
under Sec.  2.345, a petition for review under Sec.  2.341, or a motion 
for stay under Sec.  2.342, or the filing of a petition under Sec.  
2.206.
* * * * *


Sec.  2.341  [Amended]

0
19. In Sec.  2.341(a)(1), remove the phrase ``Sec.  52.103(c)'' and add 
in its place the phrase ``Sec.  52.103(c) or Sec.  53.1452(c)''.


Sec.  2.400  [Amended]

0
20. In Sec.  2.400, remove the phrase ``parts 50 or 52'' and add in its 
place the phrase ``part 50 or part 52, or Sec.  53.1470''.
0
21. In Sec.  2.401, revise the section heading and paragraph (a) to 
read as follows:


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

    (a) In the case of applications under appendix N of part 50 or 
Sec.  53.1470 of this chapter for construction permits for nuclear 
power reactors of the type described in Sec.  50.22 or part 53 of this 
chapter, or applications under appendix N of part 52 or Sec.  53.1470 
of this chapter for combined licenses, the Secretary will issue notices 
of hearing pursuant to Sec.  2.104.
* * * * *
0
22. In Sec.  2.402, revise paragraph (a) to read as follows:


Sec.  2.402  Separate hearings on separate issues; consolidation of 
proceedings.

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

[[Page 87022]]

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


Sec.  2.403  [Amended]

0
23. In Sec.  2.403, remove the phrase ``appendix N of part 50'' and add 
in its place the phrase ``appendix N to part 50 or Sec.  53.1470''.


Sec.  2.404  [Amended]

0
24. In Sec.  2.404, remove the phrase ``appendix N of part 50'' and add 
in its place the phrase ``appendix N to part 50 or Sec.  53.1470''.


Sec.  2.405  [Amended]

0
25. In Sec.  2.405, remove the phrase ``part 52'' and add in its place 
the phrase ``part 52 or part 53''.


Sec.  2.406  [Amended]

0
26. In Sec.  2.406, remove the phrase ``appendix N of parts 50 or 52'' 
and add in its place the phrase ``appendix N to part 50 or part 52 or 
Sec.  53.1470''.


Sec.  2.500  [Amended]

0
27. In Sec.  2.500, remove the phrase ``subpart F of part 52'' and add 
in its place the phrase ``subpart F of part 52 or subpart H of part 
53''.
0
28. In Sec.  2.501, revise the section heading and paragraph (a) 
introductory text to read as follows:


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

    (a) In the case of an application under subpart F of part 52 or 
subpart H of part 53 of this chapter for a license to manufacture 
nuclear power reactors of the type described in Sec.  50.22 or part 53 
of this chapter to be operated at sites not identified in the license 
application, the Secretary will issue a notice of hearing to be 
published in the Federal Register at least 30 days before the date set 
for hearing in the notice.\1\ The notice shall be issued as soon as 
practicable after the application has been docketed. The notice will 
state:
* * * * *
    \1\ The thirty-day (30) requirement of this paragraph is not 
applicable to a notice of the time and place of hearing published by 
the presiding officer after notice of hearing described in this 
section has been published.

0
29. In Sec.  2.643, revise paragraph (b) to read as follows:


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

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


Sec.  2.645  [Amended]

0
30. In Sec.  2.645, in paragraph (a), remove the phrase ``Sec.  
50.33(a) through (f) of this chapter'' and add in its place the phrase 
``Sec. Sec.  50.33(a) through (f), 53.1109, and 53.1306(a) or 53.1413 
of this chapter, as applicable,''.


Sec.  2.649  [Amended]

0
31. In Sec.  2.649, remove the phrase ``10 CFR 50.10(d)'' and add in 
its place the phrase ``10 CFR 50.10(d) or 10 CFR 53.1130(a)''.


Sec.  2.800  [Amended]

0
32. In Sec.  2.800, amend paragraphs (c) and (d) by removing the phrase 
``subpart B of part 52'' and adding in its place the phrase ``subpart B 
of part 52 or subpart H of part 53''.


Sec.  2.801  [Amended]

0
33. In Sec.  2.801, remove the phrase ``subpart B of part 52'' and add 
in its place the phrase ``subpart B of part 52 or subpart H of part 
53''.


Sec.  2.813  [Amended]

0
34. In Sec.  2.813(a), remove the phrase ``parts 50, 52, and 100'' and 
add in its place the phrase ``parts 50, 52, 53, and 100''.


Sec.  2.1103  [Amended]

0
35. In Sec.  2.1103, remove the phrase ``part 50 of this chapter'' and 
add in its place the phrase ``parts 50 or 53 of this chapter''.
0
36. In Sec.  2.1202, revise paragraphs (a)(1) through (3) and (a)(6) to 
read as follows:


Sec.  2.1202  Authority and role of NRC staff.

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


Sec.  2.1301  [Amended]

0
37. In Sec.  2.1301(b), remove ``part 50 and part 52'' and add in its 
place ``parts 50, 52, and 53''.


Sec.  2.1403  [Amended]

0
38. In Sec.  2.1403, remove the phrase ``10 CFR 50.92'' and add in its 
place the phrase ``10 CFR 50.92 or 10 CFR 53.1520''.


Sec.  2.1500  [Amended]

0
39. In Sec.  2.1500, remove the phrase ``subpart B of part 52'' and add 
in its place the phrase ``subpart B of part 52 or under subpart H of 
part 53''.


Sec.  2.1502  [Amended]

0
40. In Sec.  2.1502, in paragraph (a), remove the phrase ``Sec.  
52.51(b)'' and add in its place the phrase ``Sec. Sec.  52.51(b) or 
53.1242(b)(2)''; and in paragraph (b)(1), wherever it appears, remove 
the phrase ``Sec.  52.51(a)'' and add in its place the phrase 
``Sec. Sec.  52.51(a) or 53.1242(b)''.

PART 10--CRITERIA AND PROCEDURES FOR DETERMINING ELIGIBILITY FOR 
ACCESS TO RESTRICTED DATA OR NATIONAL SECURITY INFORMATION OR AN 
EMPLOYMENT CLEARANCE

0
41. The authority citation for part 10 continues to read as follows:

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


Sec.  10.1  [Amended]

0
42. In Sec.  10.1, in paragraph (a)(3) remove the phrase ``under part 
52'' and

[[Page 87023]]

add in its place the phrase ``under parts 52 or 53''.


Sec.  10.2  [Amended]

0
43. In Sec.  10.2, in paragraph (b), wherever it appears, remove the 
phrase ``under part 52'' and add in its place the phrase ``under parts 
52 or 53''.

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

0
44. The authority citation for part 11 continues to read as follows:

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


Sec.  11.7  [Amended]

0
45. In Sec.  11.7, in the introductory text, remove the phrase ``parts 
10, 25, 50, 70, 72, 73, and 95 of this chapter'' and add in its place 
the phrase ``parts 10, 25, 50, 53, 70, 72, 73, and 95 of this 
chapter''.

PART 19--NOTICES, INSTRUCTIONS AND REPORTS TO WORKERS: INSPECTION 
AND INVESTIGATIONS

0
46. The authority citation for part 19 continues to read as follows:

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

0
47. In Sec.  19.2, revise paragraph (a) to read as follows:


Sec.  19.2  Scope.

    (a) * * *
    (1) All persons who receive, possess, use, or transfer material 
licensed by the NRC under the regulations in parts 30 through 36, 39, 
40, 60, 61, 63, 70, or 72 of this chapter, including persons licensed 
to operate a production or utilization facility under part 50, part 52, 
or part 53 of this chapter, persons licensed to possess power reactor 
spent fuel in an independent spent fuel storage installation (ISFSI) 
under part 72 of this chapter, and in accordance with 10 CFR 76.60 to 
persons required to obtain a certificate of compliance or an approved 
compliance plan under part 76 of this chapter;
    (2) All applicants for and holders of licenses (including 
construction permits and early site permits) under parts 50, 52, 53, 
and 54 of this chapter;
    (3) All applicants for and holders of a standard design approval 
under subpart E of part 52 or under subpart H of part 53 of this 
chapter; and
    (4) All applicants for a standard design certification under 
subpart B of part 52 or under subpart H of part 53 of this chapter, and 
those (former) applicants whose designs have been certified under that 
subpart.
* * * * *
0
48. In Sec.  19.3, revise the definitions for ``License'' and 
``Regulated entities'' to read as follows:


Sec.  19.3  Definitions.

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


Sec.  19.11  [Amended]

0
49. In Sec.  19.11, in paragraph (a) introductory text, paragraph (b) 
introductory text, and paragraph (e)(1), remove the phrase ``of part 
52'' wherever may appears and add in its place the phrase ``of part 52 
or under subpart H of part 53''.


Sec.  19.14  [Amended]

0
50. In Sec.  19.14, in paragraph (a), wherever it may appear, remove 
the phrase ``of part 52'' and add in its place the phrase ``of part 52 
or under subpart H of part 53''.


Sec.  19.20  [Amended]

0
51. In Sec.  19.20, add the number ``53,'' in sequential order.

PART 20--STANDARDS FOR PROTECTION AGAINST RADIATION

0
52. The authority citation for part 20 continues to read as follows:

    Authority: Atomic Energy Act of 1954, secs. 11, 53, 63, 65, 81, 
103, 104, 161, 170H, 182, 186, 223, 234, 274, 1701 (42 U.S.C. 2014, 
2073, 2093, 2095, 2111, 2133, 2134, 2201, 2210h, 2232, 2236, 2273, 
2282, 2021, 2297f); Energy Reorganization Act of 1974, secs. 201, 
202 (42 U.S.C. 5841, 5842); Low-Level Radioactive Waste Policy 
Amendments Act of 1985, sec. 2 (42 U.S.C. 2021b); 44 U.S.C. 3504 
note.


Sec.  20.1002  [Amended]

0
53. In Sec.  20.1002, remove the phrase ``parts 30 through 36, 39, 40, 
50, 52, 60, 61, 63, 70, or 72 of this chapter'' and add in its place 
the phrase ``parts 30 through 36, 39, 40, 50, 52, 53, 60, 61, 63, 70, 
or 72 of this chapter''.
0
54. In Sec.  20.1003, revise the definition for ``License'' to read as 
follows:


Sec.  20.1003  Definitions.

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


Sec.  20.1101  [Amended]

0
55. In Sec.  20.1101, in paragraph (d), remove the phrase ``subject to 
Sec.  50.34a'' and add in its place the phrase ``subject to Sec. Sec.  
50.34a or 53.260 of this chapter''.


Sec.  20.1401  [Amended]

0
56. Amend Sec.  20.1401 by:
0
a. In paragraph (a), removing the phrase ``parts 30, 40, 50, 52, 60, 
61, 63, 70, and 72 of this chapter'', and adding in its place the 
phrase ``parts 30, 40, 50, 52, 53, 60, 61, 63, 70, and 72 of this 
chapter''; and
0
b. In paragraphs (a) and (c) removing the phrase ``in accordance with 
Sec.  50.83'' and adding in its place the phrase ``in accordance with 
Sec. Sec.  50.83 or 53.1080''.
0
57. In Sec.  20.1403, revise paragraph (d) introductory text to read as 
follows:


Sec.  20.1403  Criteria for license termination under restricted 
conditions.

* * * * *
    (d) The licensee has submitted a decommissioning plan or License 
Termination Plan (LTP) to the Commission indicating the licensee's 
intent to decommission in accordance with Sec. Sec.  30.36(d), 
40.42(d), 50.82 (a) and (b), subpart G of part 53, 70.38(d), or 72.54 
of this chapter, and specifying that the licensee intends to 
decommission by restricting use of the site. The licensee shall 
document in the LTP or decommissioning plan how the advice of 
individuals and institutions in the community who may be affected by 
the decommissioning has been sought and incorporated, as appropriate, 
following analysis of that advice.
* * * * *
0
58. In Sec.  20.1404, revise paragraph (a)(4) introductory text to read 
as follows:


Sec.  20.1404  Alternate criteria for license termination.

    (a) * * *

[[Page 87024]]

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


Sec.  20.1406  [Amended]

0
59. In Sec.  20.1406, in paragraphs (a) and (b), wherever it appears, 
remove the phrase ``under part 52'' and add in its place the phrase 
``under parts 52 or 53''.
0
60. In Sec.  20.1501, revise paragraph (b) to read as follows:


Sec.  20.1501  General.

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


Sec.  20.1905  [Amended]

0
61. In Sec.  20.1905, in paragraph (g) introductory text, remove the 
phrase ``Parts 50 or 52'' and add in its place the phrase ``parts 50, 
52, or 53''.
0
62. In Sec.  20.2004, revise paragraph (b)(1) to read as follows:


Sec.  20.2004  Treatment or disposal by incineration.

* * * * *
    (b)(1) Waste oils (petroleum derived or synthetic oils used 
principally as lubricants, coolants, hydraulic or insulating fluids, or 
metalworking oils) that have been radioactively contaminated in the 
course of the operation or maintenance of a nuclear power reactor 
licensed under parts 50 or 53 of this chapter may be incinerated on the 
site where generated provided that the total radioactive effluents from 
the facility, including the effluents from such incineration, conform 
to the requirements of appendix I to part 50 or Sec.  53.425(d) of this 
chapter and the effluent release limits contained in applicable license 
conditions other than effluent limits specifically related to 
incineration of waste oil. The licensee shall report any changes or 
additions to the information supplied under Sec. Sec.  50.34, 50.34a, 
or under subpart H of part 53 of this chapter associated with this 
incineration pursuant to Sec. Sec.  50.71 or 53.1620 of this chapter, 
as appropriate. The licensee shall also follow the procedures of 
Sec. Sec.  50.59 or 53.1565 of this chapter with respect to such 
changes to the facility or procedures.
* * * * *
0
63. In Sec.  20.2201, revise paragraphs (a)(2)(i), (b)(2)(i), and (c) 
to read as follows:


Sec.  20.2201  Reports of theft or loss of licensed material.

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


Sec.  20.2202  [Amended]

0
64. In Sec.  20.2202, in paragraph (d)(1), remove the phrase ``10 CFR 
50.72'' and add in its place the phrase ``Sec. Sec.  50.72 or 53.1630 
of this chapter;''.
0
65. In Sec.  20.2203, revise paragraph (c) to read as follows:


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

* * * * *
    (c) For holders of an operating license or a combined license for a 
nuclear power plant, the occurrences included in paragraph (a) of this 
section must be reported under the procedures described in Sec. Sec.  
50.73(b), (c), (d), (e), and (g) or 53.1640(b), (c), (d), and (e) of 
this chapter, and must include the information required by paragraph 
(b) of this section. Occurrences reported under Sec. Sec.  50.73 or 
53.1640 of this chapter need not be reported by a duplicate report 
under paragraph (a) of this section.
* * * * *


Sec.  20.2206  [Amended]

0
66. In Sec.  20.2206, in paragraph (a)(1), remove the phrase ``or Sec.  
50.22'' and add in its place the phrase ``, Sec.  50.22, or part 53''.

PART 21--REPORTING OF DEFECTS AND NONCOMPLIANCE

0
67. The authority citation for part 21 continues to read as follows:

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

0
68. In Sec.  21.2, revise paragraphs (a)(2) through (4), (b), and (c) 
to read as follows:


Sec.  21.2  Scope.

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

[[Page 87025]]

application for design certification has been granted under parts 52 or 
53 of this chapter, or who has supplied or is supplying basic 
components with respect to that design certification;
    (4) Each individual, corporation, partnership, or other entity 
doing business within the United States, and each director and 
responsible officer of such an organization, applying for or holding a 
standard design approval under parts 52 or 53 of this chapter; or 
supplying basic components with respect to a standard design approval 
under parts 52 or 53 of this chapter;
    (b) For persons licensed to construct a facility under either a 
construction permit issued under Sec. Sec.  50.23 or 53.1333 of this 
chapter or a combined license under parts 52 or 53 of this chapter (for 
the period of construction until the date that the Commission makes the 
finding under Sec. Sec.  52.103(g) or 53.1452(g) of this chapter), or 
to manufacture a facility under parts 52 or 53 of this chapter, 
evaluation of potential defects and failures to comply and reporting of 
defects and failures to comply under Sec. Sec.  50.55(e) or 53.605 of 
this chapter satisfies each person's evaluation, notification, and 
reporting obligation to report defects and failures to comply under 
this part and the responsibility of individual directors and 
responsible officers of these licensees to report defects under Section 
206 of the Energy Reorganization Act of 1974.
    (c) For persons licensed to operate a nuclear power plant under 
part 50, part 52, or part 53 of this chapter, evaluation of potential 
defects and appropriate reporting of defects under Sec. Sec.  50.72, 
50.73, 53.1630, 53.1640, or 73.1200 and 73.1205 of this chapter, 
satisfies each person's evaluation, notification, and reporting 
obligation to report defects under this part, and the responsibility of 
individual directors and responsible officers of these licensees to 
report defects under Section 206 of the Energy Reorganization Act of 
1974.
* * * * *
0
69. In Sec.  21.3, revise the definitions for ``Basic component'', 
``Commercial grade item'', ``Critical characteristics'', ``Dedicating 
entity'', ``Dedication'', ``Defect'', and ``Substantial safety hazard'' 
to read as follows:


Sec.  21.3  Definitions.

* * * * *
    Basic component. (1)(i) When applied to nuclear power plants 
licensed under part 53 of this chapter, basic component means a safety-
related structure, system, or component (SSC), or part thereof, and 
when applied to nuclear power plants licensed under parts 50 or 52, of 
this chapter, basic component means an SSC, or part thereof that 
affects its safety function necessary to assure:
    (A) The integrity of the reactor coolant pressure boundary;
    (B) The capability to shut down the reactor and maintain it in a 
safe-shutdown condition; or
    (C) The capability to prevent or mitigate the consequences of 
accidents which could result in potential offsite exposures comparable 
to those referred to in Sec. Sec.  50.34(a)(1), 50.67(b)(2), or 100.11 
of this chapter, as applicable.
    (ii) Basic components are items designed and manufactured under a 
quality assurance program complying with appendix B to part 50 of this 
chapter, or commercial grade items which have successfully completed 
the dedication process.
    (2) When applied to standard design certifications and approvals 
under part 53 of this chapter, basic component means the design or 
procurement information approved or to be approved within the scope of 
the design certification or approval for a safety-related SSC, or part 
thereof. When applied to standard design certifications under subpart B 
of part 52 of this chapter and standard design approvals under part 52 
of this chapter, basic component means the design or procurement 
information approved or to be approved within the scope of the design 
certification or approval for an SSC, or part thereof, that affects its 
safety function necessary to assure:
    (i) The integrity of the reactor coolant pressure boundary;
    (ii) The capability to shut down the reactor and maintain it in a 
safe-shutdown condition; or
    (iii) The capability to prevent or mitigate the consequences of 
accidents which could result in potential offsite exposures comparable 
to those referred to in Sec. Sec.  50.34(a)(1), 50.67(b)(2), or 100.11 
of this chapter, as applicable.
    (3) When applied to other facilities and other activities licensed 
under 10 CFR parts 30, 40, 50 (other than nuclear power plants), 60, 
61, 63, 70, 71, or 72 of this chapter, basic component means a 
structure, system, or component, or part thereof, that affects their 
safety function, that is directly procured by the licensee of a 
facility or activity subject to the regulations in this part and in 
which a defect or failure to comply with any applicable regulation in 
this chapter, order, or license issued by the Commission could create a 
substantial safety hazard.
    (4) In all cases, basic component includes safety-related design, 
analysis, inspection, testing, fabrication, replacement of parts, or 
consulting services that are associated with the component hardware, 
design certification, design approval, or information in support of an 
early site permit application under part 52 or part 53 of this chapter, 
whether these services are performed by the component supplier or 
others.
    Commercial grade item. (1) When applied to nuclear power plants 
licensed under parts 50 or 53 of this chapter, commercial grade item 
means an SSC, or part thereof that affects its safety function, that 
was not designed and manufactured as a basic component. Commercial 
grade items do not include items where the design and manufacturing 
process require in-process inspections and verifications to ensure that 
defects or failures to comply are identified and corrected (i.e., one 
or more critical characteristics of the item cannot be verified).
    (2) When applied to facilities and activities licensed pursuant to 
parts 30, 40, 50 (other than nuclear power plants), 60, 61, 63, 70, 71, 
or 72 of this chapter, commercial grade item means an item that is:
    (i) Not subject to design or specification requirements that are 
unique to those facilities or activities;
    (ii) Used in applications other than those facilities or 
activities; and
    (iii) To be ordered from the manufacturer/supplier on the basis of 
specifications set forth in the manufacturer's published product 
description (for example, a catalog).
* * * * *
    Critical characteristics. When applied to nuclear power plants 
licensed under parts 50, 52, or 53 of this chapter, critical 
characteristics are those important design, material, and performance 
characteristics of a commercial grade item that, once verified, will 
provide reasonable assurance that the item will perform its intended 
safety function.
    Dedicating entity. When applied to nuclear power plants licensed 
under parts 50, 52, or 53 of this chapter, dedicating entity means the 
organization that performs the dedication process. Dedication may be 
performed by the manufacturer of the item, a third-party dedicating 
entity, or the licensee itself. The dedicating entity, under Sec.  
21.21(c) of this part, is responsible for identifying and evaluating 
deviations, reporting defects and failures to comply for the dedicated 
item, and maintaining auditable records of the dedication process.
    Dedication. (1) When applied to nuclear power plants licensed 
pursuant to 10 CFR parts 30, 40, 50, 53, or 60, dedication is an 
acceptance process

[[Page 87026]]

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


Sec.  21.21  [Amended]

0
70. Amend Sec.  21.21 by:
0
a. In paragraph (a)(3), removing the phrase ``under part 52'' and add 
in its place the phrase ``under parts 52 or 53''; and
0
b. In paragraphs (d)(1)(i) and (ii) removing the phrase ``parts 30, 40, 
50, 52, 60, 61, 63, 70, 71, or 72 of this chapter'' and adding in its 
place the phrase ``parts 30, 40, 50, 52, 53, 60, 61, 63, 70, 71, or 72 
of this chapter''.


Sec.  21.51  [Amended]

0
71. In Sec.  21.51, in paragraphs (a)(4) and (5) remove the phrase ``of 
part 52'' and add in its place the phrase ``of part 52 or under subpart 
H of part 53''.


Sec.  21.61  [Amended]

0
72. In Sec.  21.61, in paragraph (b) remove the phrase ``under part 
52'' and add in its place the phrase ``under parts 52 or 53''.

PART 25--ACCESS AUTHORIZATION

0
73. The authority citation for part 25 continues to read as follows:

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

0
74. In Sec.  25.5, revise the definition for ``License'' to read as 
follows:


Sec.  25.5  Definitions.

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


Sec.  25.17  [Amended]

0
75. In Sec.  25.17, in paragraph (a), remove the phrase ``under 10 CFR 
parts 50, 52, 54, 60, 63, 70, 72, or 76'' and add in its place the 
phrase ``under 10 CFR parts 50, 52, 53, 54, 60, 63, 70, 72, or 76''.


Sec.  25.35  [Amended]

0
76. In Sec.  25.35, in paragraph (a), wherever it appears, remove the 
phrase ``under part 52'' and add in its place the phrase ``under parts 
52 or 53''.

PART 26--FITNESS FOR DUTY PROGRAMS

0
77. The authority citation for part 26 continues to read as follows:

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

0
78. In Sec.  26.3, revise paragraph (d) and add paragraph (f) to read 
as follows:


Sec.  26.3  Scope.

* * * * *
    (d) Contractor/vendors (C/Vs) who implement FFD programs or program 
elements, to the extent that the licensees and other entities specified 
in paragraphs (a) through (c) and (f) of this section rely on those C/V 
FFD programs or program elements to meet the requirements of this part, 
shall comply with the requirements of this part.
* * * * *
    (f) No later than the start of construction activities, licensees 
and other entities that have applied for or have been issued a license 
under part 53 of this chapter, other than a manufacturing license (ML), 
must implement the requirements in subpart M of this part or all the 
requirements of this part except subparts K and M. Holders of an ML 
under part 53 of this chapter must implement the requirements in 
subpart M or all the requirements of this part except subparts K and M, 
before commencing activities that assemble a manufactured reactor.
0
79. In Sec.  26.4, revise paragraphs (a) introductory text, (a)(1), 
(a)(4), (b), (c), (e) introductory text, (e)(4), (f), (g) introductory 
text, and (h) to read as follows:


Sec.  26.4  FFD program applicability to categories of individuals.

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

[[Page 87027]]

a licensed commercial nuclear plant, by the licensees and other 
entities in Sec.  26.3(f) and perform the following duties must be 
subject to an FFD program that satisfies the requirements in subpart M 
of this part, unless the licensee or other entity subjects these 
individuals to an FFD program that satisfies all of the requirements of 
this part except for those requirements in subparts K and M:
    (1) For persons who are granted unescorted access by the licensees 
in Sec.  26.3(a) and, as applicable, (c), operating or onsite directing 
of the operation of systems and components that a risk-informed 
evaluation process has shown to be significant to public health and 
safety; for those persons who are granted unescorted access by the 
licensees and other entities in Sec.  26.3(f), operating or directing 
of the operation of systems and components that a risk-informed 
evaluation process has shown to be significant to public health and 
safety;
* * * * *
    (4) For persons who are granted unescorted access to nuclear power 
reactor protected areas by the licensees in Sec.  26.3(a) and, as 
applicable, (c), performing maintenance or onsite directing of the 
maintenance of SSCs that a risk-informed evaluation process has shown 
to be significant to public health and safety; for those persons who 
are granted unescorted access to nuclear power reactor protected areas 
by the licensees and other entities in Sec.  26.3(f), performing 
maintenance or directing of the maintenance of SSCs that a risk-
informed evaluation process has shown to be significant to public 
health and safety; and
* * * * *
    (b) All persons who are granted unescorted access to nuclear power 
reactor protected areas by the licensees in Sec.  26.3(a) and, as 
applicable, (c) and who do not perform the duties described in 
paragraph (a) of this section shall be subject to an FFD program that 
meets all of the requirements of this part, except Sec. Sec.  26.205 
through 26.209 and subpart K of this part. All persons who are granted 
unescorted access to a facility licensed under part 53 of this chapter, 
and who do not perform or direct the performance of the duties 
described in Sec.  26.4(a), must be subject to the requirements in 
subpart M of this part, unless the licensee or other entity implements 
an FFD program that satisfies all of the requirements of this part, 
except Sec. Sec.  26.205 through 26.209 and subparts K and M.
    (c) All persons who are required by a licensee in Sec.  26.3(a) 
and, as applicable, (c) to physically report to the licensee's 
Technical Support Center or Emergency Operations Facility by licensee 
emergency plans and procedures shall be subject to an FFD program that 
meets all of the requirements of this part, except Sec. Sec.  26.205 
through 26.209 and subpart K of this part. Also, for licensees or other 
entities in Sec.  26.3(f), all persons without unescorted access to the 
facility who make decisions and/or direct actions regarding plant 
safety and security, and all persons who participate remotely in 
emergency response activities or physically report to the Technical 
Support Center or Emergency Operations Facility (or an equivalent 
facility), must be subject to an FFD program that satisfies all of the 
requirements described in subpart M of this part, unless the licensee 
or other entity implements an FFD program that satisfies all of the 
requirements of this part, except Sec. Sec.  26.205 through 26.209 and 
subparts K and M.
* * * * *
    (e) When construction activities, as defined in Sec.  26.5, begin, 
any individual whose duties for the licensees and other entities in 
Sec.  26.3(c) require him or her to have the following types of access 
or perform the following activities at the location where the nuclear 
power plant will be constructed and operated shall be subject to an FFD 
program that meets all of the requirements of this part, except 
subparts I, K, and M of this part, and for any individual whose duties 
for the licensees and other entities in Sec.  26.3(f) require him or 
her to have the following types of access, perform construction 
activities as defined in Sec.  26.5, or perform the following 
activities must be subject to an FFD program as described in subpart M 
or an FFD program that satisfies all of the requirements of this part, 
except subparts I, K, and M:
* * * * *
    (4) Witnesses or determines inspections, tests, and analyses 
certification required under part 52 or part 53 of this chapter;
* * * * *
    (f) Any individual who is constructing or directing the 
construction of safety- or security-related SSCs shall be subject to an 
FFD program that meets the requirements of subpart K, or, if 
applicable, subpart M of this part, unless the licensee or other entity 
subjects these individuals to an FFD program that meets all of the 
requirements of this part, except for subparts I, K, and M of this 
part.
    (g) All FFD program personnel who are involved in the day-to-day 
operations of the program, as defined by the procedures of the 
licensees and other entities in Sec.  26.3(a) through (c), and, as 
applicable, (d) and whose duties require them to have the following 
types of access or perform the following activities shall be subject to 
an FFD program that meets all of the requirements of this part, except 
subparts I, K, and M of this part, and, at the licensee's or other 
entity's discretion, subpart C of this part. All personnel whose duties 
require them to have the following types of access or perform the 
following activities at facilities licensed under part 53 of this 
chapter must be subject to the requirements in subpart M or an FFD 
program that satisfies all of the requirements of this part, except 
subparts I, K, and M, and, at the licensee's or other entity's 
discretion, subpart C of this part:
* * * * *
    (h) Individuals who have applied for authorization to have the 
types of access or perform the activities described in paragraphs (a) 
through (d) of this section shall be subject to Sec. Sec.  26.31(c)(1), 
26.35(b), 26.37, 26.39, and the applicable requirements of subparts C, 
E through H, and M of this part.
* * * * *
0
80. Amend Sec.  26.5 by:
0
a. Adding the definitions for ``Biological marker'' and ``Change'';
0
b. Revising the definitions for ``Constructing or construction 
activities'' ``Contractor/vendor (C/V)'';
0
c. Adding the definition of ``Illicit substance'';
0
d. Revising the definitions of ``Other entity'' and ``Questionable 
validity'';
0
e. Adding the definitions of ``Reduction in FFD program 
effectiveness'';
0
f. Revising the definitions of ``Reviewing official'', ``Safety-related 
structures, systems, and components (SSCs)'', and ``Security-related 
SSCs'';
0
g. Adding the definitions of ``Special nuclear material''; and
0
h. Revising the definition of ``Unit outage''.
    The additions and revisions read as follows:


Sec.  26.5  Definitions.

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

[[Page 87028]]

    Change as used in Sec.  26.603(e) means an action that results in a 
modification of, addition to, or removal from the licensee's or other 
entity's FFD program.
* * * * *
    Constructing or construction activities means, for the purposes of 
this part, the tasks involved in building a nuclear power plant that 
are performed at the location where the nuclear power plant will be 
constructed and operated. These tasks include fabricating, erecting, 
integrating, and testing safety- and security-related SSCs, and the 
installation of their foundations, including the placement of concrete. 
For a licensee or other entity described in Sec.  26.3(f), construction 
is defined in Sec.  53.020 of this chapter.
    Contractor/vendor (C/V) means any company, or any individual not 
employed by a licensee or other entity specified in Sec.  26.3(a) 
through (c) and (f), who is providing work or services to a licensee or 
other entity covered in Sec.  26.3(a) through (c) and (f), either by 
contract, purchase order, oral agreement, or other arrangement.
* * * * *
    Illicit substance means a substance that causes impairment and 
possible addiction but is not an illegal drug as defined in Sec.  26.5.
* * * * *
    Other entity means any corporation, firm, partnership, limited 
liability company, association, C/V, or other organization who is 
subject to this part under Sec.  26.3(a) through (c) and (f) but is not 
licensed by the NRC.
* * * * *
    Questionable validity means the results of validity screening or 
initial validity tests at a licensee testing facility indicating that a 
urine specimen may be adulterated, substituted, dilute, or invalid. For 
a part 53 licensee or other entity, questionable validity means the 
results of validity screening or initial validity tests indicating that 
a biological specimen obtained from an individual pursuant to subpart M 
of this part may be adulterated, substituted, dilute, or invalid.
    Reduction in FFD program effectiveness means, for a part 53 
licensee or other entity implementing subpart M of this part, a change 
or series of changes to an element of the FFD program that reduces or 
eliminates the licensee's ability to satisfy or maintain site-specific 
FFD program performance when compared to historical site-specific 
performance, the licensee's fleet-level program performance, or 
industry performance.
* * * * *
    Reviewing official means an employee of a licensee or other entity 
specified in Sec.  26.3(a) through (c) and (f), who is designated by 
the licensee or other entity to be responsible for reviewing and 
evaluating any potentially disqualifying FFD information about an 
individual, including, but not limited to, the results of a 
determination of fitness, as defined in Sec.  26.189, in order to 
determine whether the individual may be granted or maintain 
authorization.
    Safety-related structures, systems, and components (SSCs) means, 
for part 50 or part 52 licensees and other entities described in Sec.  
26.3(a) through (d), those SSCs that are relied on to remain functional 
during and following design basis events to ensure the integrity of the 
reactor coolant pressure boundary, the capability to shut down the 
reactor and maintain it in a safe shutdown condition, or the capability 
to prevent or mitigate the consequences of accidents that could result 
in potential offsite exposure comparable to the guidelines in Sec.  
50.34(a)(1) of this chapter. For part 53 licensees and other entities 
described in Sec.  26.3(d) and (f), safety-related has the same meaning 
as that in Sec.  53.020 of this chapter.
    Security-related SSCs means, for the purposes of this part, those 
structures, systems, and components that the licensee will rely on to 
implement the licensee's physical security and safeguards contingency 
plans that either are required under part 73 of this chapter if the 
licensee is a construction permit applicant or holder or an early site 
permit holder, as described in Sec.  26.3(c)(3) through (c)(5), 
respectively, or are included in the licensee's application if the 
licensee is a combined license applicant or holder, as described in 
Sec.  26.3(c)(1) and (c)(2), respectively, or a licensee or other 
entity described in Sec.  26.3(d) or (f).
* * * * *
    Special nuclear material (SNM) has the same meaning as that in 
Sec.  70.4 of this chapter.
* * * * *
    Unit outage means, for the purposes of this part, for electricity-
generation units, that the reactor unit is disconnected from the 
electrical grid. Unit outage means, for the purposes of this part, for 
non-electricity-generation units, that the reactor unit is disconnected 
from the loads to which its output is supplied under normal operating 
conditions.
* * * * *
0
81. In Sec.  26.8, revise paragraph (b) to read as follows:


Sec.  26.8  Information collection requirements: OMB approval.

* * * * *
    (b) The approved information collection requirements contained in 
this part appear in Sec. Sec.  26.9, 26.27, 26.29, 26.31, 26.33, 26.35, 
26.37, 26.39, 26.41, 26.53, 26.55, 26.57, 26.59, 26.61, 26.63, 26.65, 
26.67, 26.69, 26.75, 26.77, 26.85, 26.87, 26.89, 26.91, 26.93, 26.95, 
26.97, 26.99, 26.101, 26.103, 26.107, 26.109, 26.111, 26.113, 26.115, 
26.117, 26.119, 26.125, 26.127, 26.129, 26.135, 26.137, 26.139, 26.153, 
26.157, 26.159, 26.163, 26.165, 26.167, 26.168, 26.169, 26.183, 26.185, 
26.187, 26.189, 26.202, 26.203, 26.205, 26.207, 26.211, 26.401, 26.403, 
26.405, 26.406, 26.407, 26.411, 26.413, 26.415, 26.417, 26.603, 26.604, 
26.605, 26.606, 26.607, 26.608, 26.609, 26.611, 26.613, 26.617, 26.619, 
26.711, 26.713, 26.715, 26.717, 26.719, and 26.821.
0
82. Revise Sec.  26.21 to read as follows:


Sec.  26.21  Fitness-for-duty program.

    The licensees and other entities specified in Sec.  26.3(a) through 
(c) and (f) (for those licensees and other entities that do not 
implement the requirements in subparts M and K of this part) shall 
establish, implement, and maintain FFD programs that, at a minimum, 
comprise the program elements contained in this subpart. The 
individuals specified in Sec.  26.4(a) through (e) and (g), and, at the 
licensee's or other entity's discretion, Sec.  26.4(f), and, if 
necessary, Sec.  26.4(j) shall be subject to these FFD programs. 
Licensees and other entities may rely on the FFD program or program 
elements of a C/V, as defined in Sec.  26.5, if the C/V's FFD program 
or program elements satisfy the applicable requirements of this part.
0
83. Revise Sec.  26.51 to read as follows:


Sec.  26.51  Applicability.

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

[[Page 87029]]

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


Sec.  26.53  [Amended]

0
84. Amend Sec.  26.53 by:
0
a. In paragraph (e), wherever it appears, remove the phrase ``Sec.  
26.3(a) through (c)'' and add in its place the phrase ``Sec.  26.3(a) 
through (c) and (f)''; and
0
b. In paragraphs (g), (h), and (i), wherever it appears, remove the 
phrase ``(c) and (d)'' and add in its place the phrase ``(c), (d), and 
(f)''.


Sec.  26.63  [Amended]

0
85. In Sec.  26.63, in paragraph (d) remove the phrase ``Sec.  26.3(a) 
through (d)'' and add in its place the phrase ``Sec.  26.3(a) through 
(d) and (f)''.
0
86. Revise Sec.  26.73 to read as follows:


Sec.  26.73  Applicability.

    The requirements in this subpart apply to the licensees and other 
entities identified in Sec.  26.3(a), (b), and, as applicable, (c) for 
the categories of individuals specified in Sec.  26.4(a) through (d) 
and (g). The requirements in this subpart also apply to the licensees 
and other entities specified in Sec.  26.3(c), as applicable, for the 
categories of individuals in Sec.  26.4(e). At the discretion of a 
licensee or other entity in Sec.  26.3(c), the requirements of this 
subpart also may be applied to the categories of individuals identified 
in Sec.  26.4(f). In addition, the requirements in this subpart apply 
to the entities in Sec.  26.3(d) to the extent that a licensee or other 
entity relies on the C/V to satisfy the requirements of this subpart. 
The regulations in this subpart also apply to the individuals specified 
in Sec.  26.4(h) and (j), as appropriate. The requirements in this 
subpart apply to the FFD programs of licensees and other entities 
identified in Sec.  26.3(f) that elect not to implement the 
requirements in subpart M for the categories of individuals in Sec.  
26.4 and those licensees and other entities that elect to implement the 
requirements in Sec.  26.605(b).
0
87. Revise Sec.  26.81 to read as follows:


Sec.  26.81  Purpose and applicability.

    This subpart contains requirements for collecting specimens for 
drug testing and conducting alcohol tests by or on behalf of the 
licensees and other entities in Sec.  26.3(a) through (d) for the 
categories of individuals specified in Sec.  26.4(a) through (d) and 
(g). At the discretion of a licensee or other entity in Sec.  26.3(c), 
specimen collections and alcohol tests must be conducted either under 
this subpart for the individuals specified in Sec.  26.4(e) and (f) or 
the licensee or other entity may rely on specimen collections and 
alcohol tests conducted under the requirements of 49 CFR part 40 for 
the individuals specified in Sec.  26.4(e) and (f). The requirements of 
this subpart do not apply to specimen collections and alcohol tests 
that are conducted under the requirements of 49 CFR part 40, as 
permitted in this paragraph and under Sec. Sec.  26.4(j) and 
26.31(b)(2) and subpart K. The requirements in this subpart apply to 
the FFD programs of licensees and other entities identified in Sec.  
26.3(f) that elect not to implement the requirements in subpart M for 
the categories of individuals in Sec.  26.4 and those licensees and 
other entities that elect to implement the requirements in Sec.  
26.605.
0
88. Revise Sec.  26.201 to read as follows:


Sec.  26.201  Applicability.

    (a) The requirements in this subpart, with the exception of Sec.  
26.202, apply to the licensees and other entities identified in Sec.  
26.3(a); if applicable, (c), (d), and (f), for licensees and other 
entities not implementing the requirements in subparts K and M. For the 
licensees and other entities to whom the requirements in this subpart, 
with the exception of Sec.  26.202, apply, the requirements in 
Sec. Sec.  26.203 and 26.211 apply to the individuals identified in 
Sec.  26.4(a) through (c). In addition, the requirements in Sec. Sec.  
26.205 through 26.209 apply to the individuals identified in Sec.  
26.4(a).
    (b) The requirements in this subpart, with the exception of Sec.  
26.203, apply to the licensees or other entities identified in Sec.  
26.3(f) implementing this subpart under Sec. Sec.  26.604 and 26.605. 
For these licensees and other entities, the requirements in Sec. Sec.  
26.202 and 26.211 apply to the individuals identified in Sec.  26.4(a) 
through (c) and any person licensed to operate under 10 CFR part 53; 
and the requirements in Sec. Sec.  26.205 through 26.209 apply to the 
individuals identified in Sec.  26.4(a).
0
89. Add Sec.  26.202 to read as follows:


Sec.  26.202  General provisions for facilities licensed under part 53.

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

[[Page 87030]]

who are subject to the work hour controls in Sec.  26.205;
    (3) The documentation of waivers that is required in Sec.  
26.207(a)(4), including the bases for granting the waivers;
    (4) The documentation of work hour reviews that is required in 
Sec.  26.205(e)(3) and (e)(4); and
    (5) The documentation of fatigue assessments that is required in 
Sec.  26.211(g).
    (e) Reporting. Licensees must include the following information in 
a standard format in the annual FFD program performance report required 
under Sec.  26.617(b)(2):
    (1) A summary for each nuclear power plant site of all instances 
during the previous calendar year when the licensee waived one or more 
of the work hour controls specified in Sec.  26.205(d)(1) through 
(d)(5)(i) and (d)(7) for individuals described in Sec.  26.4(a). The 
summary must include only those waivers under which work was performed. 
If it was necessary to waive more than one work hour control during any 
single extended work period, the summary of instances must include each 
of the work hour controls that were waived during the period. For each 
category of individuals specified in Sec.  26.4(a), the licensee must 
report--
    (i) The number of instances when each applicable work hour control 
specified in Sec.  26.205(d)(1)(i) through (iii), (d)(2)(i) and (ii), 
(d)(3)(i) through (v), and (d)(7) was waived for individuals not 
working on outage activities;
    (ii) The number of instances when each applicable work hour control 
specified in Sec.  26.205(d)(1)(i) through (iii), (d)(2)(i) and (ii), 
(d)(3)(i) through (v), (d)(4) and (d)(5)(i), and (d)(7) was waived for 
individuals working on outage activities; and
    (iii) A summary that shows the distribution of waiver use among the 
individuals applicable within each category of individuals identified 
in Sec.  26.4(a) (e.g., a table that shows the number of individuals 
who received only one waiver during the reporting period, the number of 
individuals who received a total of two waivers during the reporting 
period).
    (2) A summary of corrective actions, if any, resulting from the 
analyses of these data, including fatigue assessments.
    (f) Audits. Licensees must audit the management of worker fatigue 
under Sec.  26.615.
0
90. In Sec.  26.205, revise paragraphs (d)(7)(iii) and (d)(8) to read 
as follows:


Sec.  26.205  Work Hours.

* * * * *
    (d) * * *
    (7) * * *
    (iii) Each licensee shall state, in its FFD policy and procedures 
required by either Sec. Sec.  26.27 and 26.203(a) and (b) or Sec. Sec.  
26.202(a) and (b) and 26.606, the work hour counting system in Sec.  
26.205(d)(7)(ii) the licensee is using.
    (8) Each licensee shall state, in its FFD policy and procedures 
required by either Sec. Sec.  26.27 and 26.203(a) and (b) or Sec. Sec.  
26.202(a) and (b) and 26.606, the requirements with which the licensee 
is complying: the minimum days off requirements in Sec.  26.205(d)(3) 
or maximum average work hours requirements in Sec.  26.205(d)(7).
* * * * *
0
91. In Sec.  26.207, revise paragraph (a)(1)(ii) to read as follows:


Sec.  26.207  Waivers and exceptions.

    (a) * * *
    (1) * * *
    (ii) A supervisor assesses the individual face to face and 
determines that there is reasonable assurance that the individual will 
be able to safely and competently perform his or her duties during the 
additional work period for which the waiver will be granted. The 
supervisor performing the assessment shall be trained as required by 
either Sec. Sec.  26.29 and 26.203(c) or Sec. Sec.  26.202(c) and 
26.608 and shall be qualified to direct the work to be performed by the 
individual. If there is no supervisor on site who is qualified to 
direct the work, the assessment may be performed by a supervisor who is 
qualified to provide oversight of the work to be performed by the 
individual. At a minimum, the assessment must address the potential for 
acute and cumulative fatigue considering the individual's work history 
for at least the past 14 days, the potential for circadian degradations 
in alertness and performance considering the time of day for which the 
waiver will be granted, the potential for fatigue-related degradations 
in alertness and performance to affect risk-significant functions, and 
whether any controls and conditions must be established under which the 
individual will be permitted to perform work. For licensees and other 
entities in Sec.  26.3(f), the assessment may be performed remotely 
using electronic communications. In such instances, the assessment must 
be supported by someone who is present in-person with the individual 
whose alertness may be impaired, and that supporting person must be 
trained under the requirements of either Sec.  26.29 and Sec.  
26.203(c) or Sec.  26.202(c) and Sec.  26.608.
* * * * *
0
92. In Sec.  26.211, revise paragraphs (a)(1) and (3) and paragraph (b) 
introductory text to read as follows:


Sec.  26.211  Fatigue assessments.

    (a) * * *
    (1) For-cause. In addition to any other test or determination of 
fitness that may be required under Sec. Sec.  26.31(c), 26.77, 
26.607(b), and 26.619, a fatigue assessment must be conducted in 
response to an observed condition of impaired individual alertness 
creating a reasonable suspicion that an individual is not fit to safely 
and competently perform his or her duties, except if the condition is 
observed during an individual's break period. If the observed condition 
is impaired alertness with no other behaviors or physical conditions 
creating a reasonable suspicion of possible substance abuse, then the 
licensee need only conduct a fatigue assessment. If the licensee has 
reason to believe that the observed condition is not due to fatigue, 
the licensee need not conduct a fatigue assessment;
* * * * *
    (3) Post-event. A fatigue assessment must be conducted in response 
to events requiring post-event drug and alcohol testing as specified in 
Sec.  26.31(c) or post-event tests in Sec.  26.607(b)(4). Licensees may 
not delay necessary medical treatment in order to conduct a fatigue 
assessment; and
* * * * *
    (b) Only supervisors and FFD program personnel who are trained 
under either Sec. Sec.  26.29 and 26.203(c) or Sec. Sec.  26.202(c) and 
26.608 may conduct a fatigue assessment. The fatigue assessment must be 
conducted face to face with the individual whose alertness may be 
impaired. For licensees and other entities in Sec.  26.3(f), a fatigue 
assessment may be performed remotely using electronic communications. 
In such instances, the fatigue assessment must be supported by someone 
who is present in-person with the individual whose alertness may be 
impaired, and that supporting person must be trained in accordance with 
the requirements of either Sec. Sec.  26.29 and 26.203(c) or Sec. Sec.  
26.202(c) and 26.608.
* * * * *
0
93. Add Subpart M, consisting of Sec. Sec.  26.601 through 26.619, to 
read as follows:

Subpart M--Fitness for Duty Programs for Facilities Licensed Under 
10 CFR Part 53

Sec.
26.601 Applicability.
26.603 General provisions.

[[Page 87031]]

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


Sec.  26.601  Applicability.

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


Sec.  26.603  General provisions.

    (a) FFD program description. An applicant's description of the FFD 
program in its Final Safety Analysis Report, required by subpart H of 
part 53 of this chapter, must include--
    (1) If the applicant performed the analysis under paragraph (c) of 
this section, a summary of the analysis, including the assumptions, 
methodology, conclusion, and references;
    (2) A statement whether the FFD program will be implemented 
pursuant to Sec.  26.604 or Sec.  26.605, or will satisfy all part 26 
requirements, except for the requirements in subparts K and M;
    (3) A discussion of the applicability of the FFD program to those 
individuals described in Sec.  26.4 and how the program will be 
implemented offsite at a U.S. Nuclear Regulatory Commission (NRC)-
licensed facility authorized to assemble or test a manufactured 
reactor, if applicable;
    (4) A description of the drug and alcohol testing and fitness 
determination process to be implemented through the licensee's or other 
entity's procedures, including the collection and testing facilities to 
be used, biological specimens to be collected, and sanctions to be 
imposed upon a confirmed FFD policy violation; and
    (5) A summary of the FFD performance monitoring and review program 
(PMRP), including the measures and thresholds required by paragraph 
(d)(1) of this section.
    (b) FFD program implementation and availability. For the licensees 
and other entities in Sec.  26.3(f), other than the holder of a 
manufacturing license (ML), the FFD program must be implemented no 
later than the start of construction activities, as defined in Sec.  
26.5, and maintained until the NRC's docketing of the license holder's 
certifications described in Sec.  53.1070 of this chapter. For holders 
of an ML, the FFD program must be implemented no later than the start 
of activities that assemble the manufactured reactor and maintained 
until expiration of the ML.
    (c) Criterion and analysis for an FFD program. For a licensee or 
other entity to implement an FFD program under Sec.  26.604, the 
licensee or other entity must perform a site-specific analysis to 
demonstrate that the facility and its operation satisfy the criterion 
in Sec.  53.860(a)(2) of this chapter. The licensee or other entity 
must maintain the analysis, including updates to reflect changes made 
to the staffing, FFD programs, or offsite support resources described 
in the analysis, to show that the facility and its operation continue 
to satisfy the criterion, until permanent cessation of operations under 
Sec.  53.1070 of this chapter.
    (d) FFD performance monitoring and review. A licensee or other 
entity must establish performance measures and associated thresholds as 
described in paragraph (d)(1) of this section and monitor the 
effectiveness of its FFD program by comparing performance data against 
these performance measures and thresholds, in a manner sufficient to 
satisfy the Sec.  26.23 performance objectives.
    (1) PRMP elements. The PMRP must be documented and maintained and 
include the following program elements:
    (i) Performance measures. Performance measures must be identified 
and designed to monitor FFD program performance.
    (A) If the licensee or other entity is subject to the requirements 
in Sec.  26.604, then the monitoring program must include performance 
measures for the following: the behavioral observation program; 
occurrence of FFD policy violations categorized by licensee employee, 
contractor/vendor, and labor category; and occurrence of individuals 
with potentially disqualifying information or who possessed FFD 
prohibited items.
    (B) If the licensee or other entity is subject to the requirements 
in Sec.  26.604 and has implemented a drug testing program at its 
discretion or is subject to the requirements of Sec.  26.605, then the 
monitoring program must include performance measures identified in 
paragraph (d)(1)(i)(A) of this section. This monitoring program must 
also include performance measures for the pre-access and random 
positive testing rates, random testing rate for licensee employees and 
contractor/vendors, and the number of subversion attempts categorized 
by licensee employee, contractor/vendor, and labor category.
    (ii) Thresholds. Licensee- or other entity-specific thresholds for 
its site-specific performance measures must be established and used to 
facilitate corrective actions to maintain FFD program performance. 
Initial thresholds must be based on FFD performance data from 
comparable facilities subject to part 26, the licensee's or other 
entity's fleet-level program performance if applicable, and industry 
FFD performance data.
    (iii) Monitoring program. Licensees and other entities must monitor 
the performance of their FFD programs against licensee- or other 
entity-established performance measures and thresholds as FFD 
performance data is received to determine whether a threshold has been 
exceeded. Licensees and other entities must perform year-to-year 
comparisons of site-specific performance; site-specific performance to 
the licensee's or other entity's fleet-level program performance, if 
applicable; and site-specific to industry performance.
    (iv) Quantitative and qualitative reviews. The PMRP must include a 
documented review of the elements in paragraph (d)(1)(i) through (iii) 
of this section and the following qualitative elements.
    (A) Worker protections. The review must include a documented 
assessment of the licensee's or other entity's implementation of the 
protections described in Sec. Sec.  26.606(b)(1)(iii), 26.611, and 
26.613.
    (B) Laboratory test results and Medical Review Officer performance. 
The review must include a documented assessment of whether the actions 
taken by the Medical Review Officer (MRO) met the requirements in Sec.  
26.185 based on the laboratory test results reported under Sec.  
26.169. This review must include a comparative analysis between the 
point of collection testing and assessment (POCTA) screening result(s) 
and the corresponding specimen test results obtained from the U.S. 
Department of Health and Human

[[Page 87032]]

Services (HHS)-certified laboratory if the POCTA indicated a positive, 
adulterated, substituted, or invalid screening result or discrepant 
biological marker, to assess the effectiveness of the POCTA and to 
inform MRO decisions under Sec.  26.185 or Sec.  26.607(m)(6).
    (C) Change control. The review must include a documented assessment 
of the changes made under paragraph (e) of this section to verify that 
the summation of program changes has not resulted in a reduction in FFD 
program effectiveness.
    (2) Corrective actions. Corrective actions must be implemented to 
address when FFD performance meets a licensee-established performance 
threshold or to resolve a finding resulting from a qualitative review 
or audit in a manner that restores performance and corrects root 
causes, contributing causes, or both.
    (3) Program review periodicity. The documented review in paragraph 
(d)(1)(iv) of this section must be conducted biennially to assess and 
modify licensee or other entity implementation of its FFD program. This 
documented review must demonstrate that the performance measures and 
thresholds are appropriate and adjusted as necessary based on site-
level and licensee's or other entity's fleet-level, if applicable, 
program performance, and industry performance.
    (i) Identified program weaknesses and corrective actions must be 
summarized in the annual reporting requirement described in Sec.  
26.617(b)(2) or Sec.  26.717, as applicable.
    (ii) The program review must be completed and approved by the 
licensee or other entity to support the reporting of PMRP weaknesses 
and corrective actions as required in paragraph (d)(3)(i) of this 
section every odd-numbered year, and the implementation of corrective 
actions before May 15 of that odd-numbered year.
    (e) FFD program change control. (1) The licensee or other entity 
may make changes to its FFD program under this subpart if--
    (i) The licensee or other entity performs and retains an analysis 
demonstrating that the changes do not reduce the effectiveness of the 
FFD program; or
    (ii) The change was necessitated or justified by a change to part 
26, laboratory processes or procedures, or guidance issued by the HHS 
or NRC, as implemented by the licensee or other entity though its 
procedures.
    (2) A licensee or other entity desiring to make a change that 
decreases FFD program effectiveness must implement a mitigating 
strategy so the FFD program, as revised, will continue to satisfy the 
performance objectives in Sec.  26.23 and not result in a reduction in 
program effectiveness.
    (3) Except for phencyclidine, and notwithstanding paragraph 
(e)(1)(ii) of this section, the change control process may not be used 
to reduce the minimum panel of drugs to be tested in Sec.  
26.607(c)(1).
    (4) The licensee must retain a record of each change made under 
this section for a period of at least 5 years from the date the change 
was implemented and summarize this change in its annual FFD performance 
report required by Sec.  26.617(b)(2) or Sec.  26.717, as applicable.


Sec.  26.604  FFD program requirements for facilities that satisfy the 
Sec.  26.603(c) criterion.

    (a) FFD program. A licensee or other entity with an analysis that 
demonstrates that its facility and operation satisfy the criterion in 
Sec.  26.603(c) may elect to establish, implement, and maintain an FFD 
program under this section. That FFD program must contain the following 
elements:
    (1) Applies to those individuals described in Sec.  26.4, as 
applicable; and
    (2) Implements the following requirements and subparts in this 
part:
    (i) Sec.  26.23, Performance objectives;
    (ii) Sec.  26.603, General provisions;
    (iii) Sec.  26.606, Written policies and procedures, (a) and, if 
applicable (b);
    (iv) Sec.  26.608, FFD program training;
    (v) Sec.  26.609, Behavioral observation;
    (vi) Sec.  26.610, Sanctions;
    (vii) Sec.  26.611, Protection of information;
    (viii) Sec.  26.613, Appeals process;
    (ix) Sec.  26.615, Audits;
    (x) Sec.  26.617, Recordkeeping and reporting;
    (xi) Sec.  26.619, Suitability and fitness determinations;
    (xii) Subpart A--Administrative Provisions;
    (xiii) Subpart I--Managing Fatigue; and
    (xiv) Subpart O--Inspections, Violations, and Penalties.
    (b) [Reserved]


Sec.  26.605  FFD program requirements for facilities that do not 
implement Sec.  26.604.

    (a) Licensees and other entities who satisfy the criterion in Sec.  
26.603(c), at their discretion, and licensees and other entities who do 
not satisfy the criterion in Sec.  26.603(c), must establish, 
implement, and maintain an FFD program under this section either during 
construction activities as defined in Sec.  26.5, or during activities 
performed under an ML that allows the assembly, testing, or both of a 
manufactured reactor, as applicable. This FFD program must contain the 
following elements:
    (1) Applies to those individuals described in Sec.  26.4, as 
applicable; and,
    (2) Implements the following requirements and subparts in this 
part--
    (i) Sec.  26.23, Performance objectives;
    (ii) Sec.  26.603, General provisions;
    (iii) Sec.  26.606, Written policy and procedures;
    (iv) Sec.  26.607, Drug and alcohol testing;
    (v) Sec.  26.608, FFD program training;
    (vi) Sec.  26.609, Behavioral observation;
    (vii) Sec.  26.610, Sanctions;
    (viii) Sec.  26.611, Protection of information;
    (ix) Sec.  26.613, Appeals process;
    (x) Sec.  26.615, Audits;
    (xi) Sec.  26.617, Recordkeeping and reporting;
    (xii) Sec.  26.619, Suitability and fitness determinations;
    (xiii) Subpart A--Administrative Provisions;
    (xiv) Subpart I--Managing Fatigue, in the case of holders of an ML 
that allows the assembly, testing, or both of a manufactured reactor; 
and
    (xv) Subpart O--Inspections, Violations, and Penalties.
    (b) Licensees and other entities who satisfy the criterion in Sec.  
26.603(c), at their discretion, and licensees and other entities who do 
not satisfy the criterion in Sec.  26.603(c), before the loading of 
fuel onsite into a reactor vessel; before receiving a manufactured 
reactor; or before individuals subject to part 26 operate, test, 
perform maintenance of, or direct the maintenance or surveillance of 
security-related equipment or equipment that a risk-informed evaluation 
process has shown to be significant to public health and safety, must 
establish, implement, and maintain an FFD program that--
    (1) Applies to those individuals described in Sec.  26.4, as 
applicable; and,
    (2) Implements the following requirements and subparts--
    (i) Sec.  26.23, Performance objectives;
    (ii) Sec.  26.603, General provisions;
    (iii) Sec.  26.606, Written policy and procedures;
    (iv) Sec.  26.607, Drug and alcohol testing;
    (v) Sec.  26.608, FFD program training;
    (vi) Sec.  26.609, Behavioral observation;
    (vii) Sec.  26.611, Protection of information;
    (viii) Sec.  26.613, Appeals process;
    (ix) Sec.  26.615, Audits;
    (x) Subpart A--Administrative Provisions;
    (xi) Subpart C--Granting and Maintaining Authorization;
    (xii) Subpart D--Management Actions and Sanctions to be Imposed;
    (xiii) Subpart H--Determining Fitness-for-Duty Policy Violations 
and

[[Page 87033]]

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


Sec.  26.606  Written policy and procedures.

    (a) Licensees and other entities that implement an FFD program 
under this subpart must ensure that--
    (1) A written FFD policy statement is provided to each individual 
who is subject to the program before the individual is subject to 
behavioral observation, drug and alcohol testing, or both.
    (2) The FFD policy statement describes the performance objectives 
in Sec.  26.23.
    (3) The FFD policy statement describes the minimum days off 
requirements in Sec.  26.205(d)(3) or maximum average work hours 
requirements in Sec.  26.205(d)(7).
    (4) The FFD policy statement must be written in sufficient detail 
to provide affected individuals with information on what is expected of 
them and what consequences may result from a lack of adherence to the 
policy, including those elements described in Sec.  26.606(b), part 26-
required sanctions, and required medical/clinical treatment and follow-
up testing for FFD policy violations.
    (5) The FFD policy statement describes the individual's 
responsibilities to report for work in a physiological and 
psychological condition that enables the safe and competent performance 
of assigned duties and responsibilities and inform a licensee- or other 
entity-designated representative when the individual determines that 
this cannot be accomplished.
    (b) Licensees and other entities must establish, implement, and 
maintain written procedures that address the following topics:
    (1) If implementing a drug and alcohol testing program under this 
subpart,
    (i) The methods and techniques to collect and test for drugs and 
alcohol and for the shipping and temporary storage of biological 
specimens used for drug testing at HHS-certified laboratories,
    (ii) The urine specimen volumes, techniques for split specimen 
collections, and the acceptability of a urine specimen as described in 
Sec.  26.111 or as described in the HHS Guidelines,
    (iii) Protecting the privacy of an individual who provides a 
specimen, protecting the integrity of the specimen, and ensuring that 
the test results are valid and attributable to the correct individual, 
and
    (iv) If the licensee or other entity elects to use the HHS 
Guidelines, the name of the specific HHS Guideline and revision being 
implemented by the licensee or other entity and a description of the 
specific sections in the guideline that are being implemented in the 
procedure, including specimen collections, drug testing, and evaluation 
of test results.
    (2) The immediate and follow-up actions that will be taken, and the 
procedures to be used, in those cases in which individuals who are 
subject to the FFD program:
    (i) Have been involved in the use, sale, or possession of illegal 
substances, illegal drugs, or illicit substances;
    (ii) Are impaired by any illegal substances, illegal drugs, or 
illicit substances or the consumption of alcohol as determined by 
behavioral observation or a test that measures blood alcohol 
concentration;
    (iii) If drug and alcohol testing is conducted, attempted to 
subvert the testing process by adulterating or diluting specimens (in 
vivo or in vitro), substituting specimens, or by any other means;
    (iv) If drug and alcohol testing is conducted, refused to provide a 
specimen for analysis or follow instructions provided by FFD program 
personnel;
    (v) Had legal action taken relating to drug or alcohol use; or
    (vi) Demonstrated character or actions indicating that the 
individual cannot be trusted or relied upon to perform those duties and 
responsibilities or maintain access to NRC-licensed facilities, special 
nuclear material (SNM), or sensitive information.
    (3) The process, including the duties and responsibilities of FFD 
program personnel, to be followed if an individual's behavior or 
condition raises a concern regarding the possible use, sale, or 
possession of illegal drugs on- or offsite; the possible use or 
possession of alcohol on the NRC-licensed facility; impairment from any 
cause that in any way could adversely affect the individual's ability 
to safely and competently perform the individual's duties; or the 
receipt of credible information indicating that the individual cannot 
be trusted or relied on to perform those duties and responsibilities 
making the individual subject to this part.
    (4) Operation and oversight of an onsite or offsite collection 
facility.
    (5) The fatigue management requirements in Sec. Sec.  26.202(b) and 
either 26.205(d)(3) or (d)(7).
    (6) Measures to prevent subversion of drug and alcohol tests 
conducted onsite and offsite.


Sec.  26.607  Drug and alcohol testing.

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

[[Page 87034]]

testing provides that the number of random tests performed annually is 
equal to at least 50 percent for licensee employees and 50 percent for 
contractor/vendors at the NRC-licensed site.
    (3) For-cause. A for-cause drug test, alcohol test, or both, must 
be conducted onsite in response to an individual's observed behavior or 
physical condition indicating possible substance abuse or after 
receiving credible information that an individual is engaging in 
substance abuse, as defined in Sec.  26.5;
    (4) Post-event. A post-event test for drugs and alcohol must be 
conducted--
    (i) As soon as practical after an event involving a human error 
that was committed by an individual specified in Sec.  26.4, where the 
human error may have caused or contributed to the event. This test must 
be conducted onsite unless the individual requires offsite medical 
care. The licensee or other entity must test the individual(s) who 
committed or directed the error and need not test individuals who were 
affected by the event and whose actions likely did not cause or 
contribute to the event. The licensee or other entity must describe in 
its procedures what constitutes a human error.
    (ii) Within 4 hours of an event unless immediate medical 
intervention precludes the conduct of the test on the individual(s) who 
caused or contributed to the accident(s), if the event results in--
    (A) An illness or personal injury to any individual which results 
in death, days away from work, restricted work, transfer to another 
job, medical treatment beyond first aid, loss of consciousness, or 
other significant illness or injury, as diagnosed by a licensee- or 
other entity-designated physician or other licensed health care 
professional, even if the illness or injury does not result in death, 
days away from work, restricted work or job transfer, medical treatment 
beyond first aid, or loss of consciousness; or
    (B) Damage to any safety- or security-related structures, systems, 
and components; and
    (5) Follow-up. An individual subject to part 26 who has violated 
the FFD policy for substance use or abuse, or the sale, use, or 
possession of illegal drugs must be subject to a follow-up series of 
tests for drugs, alcohol, or both to verify an individual's continued 
abstinence from substance abuse.
    (c) Urine and oral fluid specimens. (1) All urine or oral fluid 
specimens must be subject to validity testing, including an adulterant 
and biological marker, and tested for the substances listed in Sec.  
26.31(d)(1), except as allowed by Sec.  26.603(e)(3).
    (2) For the use of urine as the biological specimen to be tested, 
the following requirements must be implemented--
    (i) Sec.  26.115, Collecting a urine specimen under direct 
observation;
    (ii) Sec.  26.119, Determining ``shy'' bladder; and
    (iii) Sec.  26.163, Cutoff levels for drugs and drug metabolites, 
(a)(2) regarding special analysis testing.
    (3) For alcohol testing onsite, the following requirements must be 
implemented--
    (i) Sec.  26.91, Acceptable devices for conducting initial and 
confirmatory tests for alcohol and methods of use;
    (ii) Sec.  26.93, Preparing for alcohol testing;
    (iii) Sec.  26.95, Conducting an initial test for alcohol using a 
breath specimen;
    (iv) Sec.  26.97, Collecting oral fluid specimens for alcohol and 
drug testing;
    (v) Sec.  26.99, Determining the need for a confirmatory test for 
alcohol;
    (vi) Sec.  26.101, Conducting a confirmatory test for alcohol; and,
    (vii) Sec.  26.103, Determining a confirmed positive test result 
for alcohol.
    (4) For all test conditions in paragraph (b) of this section, 
except for the use of a POCTA screening device in paragraph (h) of this 
section, and for MRO-directed tests under Sec.  26.185, drug testing 
must be performed at an HHS-certified laboratory for the specific 
biological specimen to be tested. Only HHS-certified laboratory test 
results from urine and oral fluid specimens may be used for the 
issuance of a part 26-required sanction. The licensee or other entity 
must establish and maintain a contract with a primary and a back-up 
HHS-certified laboratory (with a different Certifying Scientist) for 
the specimen(s) to be tested. These contracts must stipulate that the 
laboratories are subject to inspection or audit by the licensee or 
other entity and that records and documents must be provided and/or 
able to be photocopied and removed from the premises to support the 
inspection or audit.
    (d) Privacy and integrity. The specimen collection and drug and 
alcohol testing procedures of FFD programs must protect the donor's 
privacy and the integrity of the specimen and implement quality 
controls to ensure that test results are valid and attributable to the 
correct individual.
    (e) Offsite collection facilities. At the licensee's or other 
entity's discretion, specimen collections and alcohol testing may be 
conducted at a local hospital or other facility licensed to conduct 
specimen collections and perform alcohol testing and audited by the 
State or a State-designated entity. The licensee or other entity must 
audit these facilities, if used, before their initial use and then on a 
biennial basis to confirm that the facility procedures are comparable 
to those described in subpart E of this part or the HHS Guidelines for 
urine and oral fluid.
    (f) Initial testing. A licensee or other entity subject to this 
subpart performing an initial test must use an immunoassay, or an 
alternative technology established in its FFD program through Sec.  
26.603(e), that satisfies the requirements of the U.S. Food and Drug 
Administration (FDA) for commercial distribution. Specimens that yield 
positive, adulterated, substituted, or invalid initial validity or drug 
test results or discrepant biological markers must be subject to 
confirmatory testing by an HHS-certified laboratory, certified for that 
biological specimen, except for invalid specimens that cannot be 
tested.
    (g) Oral fluid testing. If the licensee or other entity elects to 
use oral fluid for drug or alcohol testing, the collection, packaging, 
and temporary storage of the drug or alcohol test device, and shipment 
of an oral fluid specimen to an HHS-certified laboratory or the 
collection of an oral fluid specimen for alcohol testing must be 
performed in accordance with licensee- or other entity-established 
procedures based either on the requirements in part 26 or the 
procedures in HHS Guidelines identified by the licensee or other entity 
in Sec.  26.606(b)(1)(iv). The device must have received premarket 
approval from the FDA and must not expire before laboratory testing. 
The drugs, drug metabolites, initial and confirmatory testing cutoffs, 
and biological markers, if applicable, must be those established by HHS 
for oral fluid testing and the alcohol cutoffs in this part or, if not 
established by HHS or the NRC for the panel of drugs and drug 
metabolites to be tested, as determined and documented by a forensic 
toxicologist review conducted pursuant to Sec.  26.31(d)(1)(i)(D).
    (h) Point of collection testing and assessment. (1) If the licensee 
or other entity elects to use a POCTA device, then it may only be used 
for pre-access and random drug and alcohol initial testing in paragraph 
(b) of this section, the alcohol testing process in paragraph (c)(3) of 
this section, and the portal area screening process in paragraph (j) of 
this section. Before the licensee or other entity uses a POCTA device, 
a forensic toxicologist must review and document

[[Page 87035]]

their evaluation that the validity and accuracy of the device for 
alcohol and/or the drugs and drug metabolites listed in Sec.  26.31(d) 
are comparable to the performance achieved by initial testing conducted 
using a similar technology at an HHS-certified laboratory. For initial 
testing of drugs and drug metabolites using a POCTA device, this review 
must include a documented evaluation of POCTA device performance 
against the requirements in Sec.  26.161(b) for a urine specimen or the 
procedures in the HHS Guidelines for urine or oral fluid, as 
implemented by the licensee or other entity through its procedures.
    (2) If the performance of the POCTA device is not comparable to 
that achieved from initial testing conducted by an HHS-certified 
laboratory as determined by the forensic toxicologist, then the 
licensee or other entity must implement a mitigating strategy to 
maintain program effectiveness under Sec.  26.603(e)(2), as applicable.
    (3) The licensee and other entity must implement procedures for the 
use of a POCTA that ensures the effectiveness of the collection 
process, assessment of the screening results, and prevention of 
subversion attempts.
    (4) If the use of a POCTA device indicates a discrepant biological 
marker or that a test result exceeds the initial test cutoff, the 
specimen is invalid, or the individual subverted the drug or alcohol 
test, then the individual must be immediately removed from duties, 
responsibilities, and access making the individual subject to this 
subpart.
    (i) The individual must be subject to an immediate drug and alcohol 
test using the alcohol testing process in paragraph (c)(3) of this 
section for a positive alcohol screen and either oral fluid or urine by 
a collection kit that is not a POCTA device, but of the same type of 
biological specimen collected by the POCTA, for validity, if required, 
and initial and confirmatory testing by an HHS-certified laboratory.
    (ii) If this individual shows any signs of impairment, the 
individual's authorization must be temporarily removed until the MRO 
reviews the laboratory test result(s), interviews the individual, and 
performs a determination of fitness under Sec.  26.189 or Sec.  26.619, 
as applicable, that enables the restoration of authorization.
    (i) Hair testing. The testing of hair specimens may only be used to 
inform a licensee's or other entity's determination of whether the 
individual is trustworthy and reliable under the test condition in 
paragraph (b)(1) of this section to supplement the information gained 
from a pre-access test using oral fluid or urine as the test specimen 
and must be conducted at an HHS-certified laboratory certified for hair 
specimens.
    (1) If used, this process must be described in the licensee's or 
other entity's FFD policy and described in detail in its procedure. The 
panel of drugs and drug metabolites to be evaluated must only include 
those listed as Schedule I or II of section 202 of the Controlled 
Substances Act [21 U.S.C. 812]. The collection, packaging, and 
temporary storage of a hair specimen and shipment of the specimen to an 
HHS-certified laboratory must be conducted in accordance with the HHS 
Guidelines. The test kit must be FDA cleared, and licensee- or other 
entity-designated FFD program personnel must conduct the collection, 
packaging, temporary storage, shipping, and custody and control of the 
specimen.
    (2) Before the licensee or other entity begins to conduct hair 
testing, the initial and confirmatory testing cutoffs must be the 
cutoffs established by HHS for hair testing or, if not established by 
HHS or the NRC, as determined by a forensic toxicologist review 
conducted pursuant to Sec.  26.31(d)(1)(i)(D).
    (3) Confirmed positive test results must be considered potentially 
disqualifying FFD information until proven otherwise by a review under 
Sec.  26.613. Sanctions under this subpart must not be issued for any 
FFD policy violation involving a drug test using a hair specimen unless 
the licensee or other entity determines that the individual subverted, 
as defined in Sec.  26.5, the hair test.
    (j) Portal area screening. A non-invasive point of collection 
testing instrument may be used to screen individuals for drugs, drug 
metabolites, and alcohol before the individuals' entry into or exit 
from a protected or vital area.
    (1) If a licensee or other entity uses such an instrument, then 
before such use, a forensic toxicologist must review the instrument and 
document an evaluation that the instrument and setpoints used in the 
instrument are acceptable for use for the detection and screening of 
the drugs and drug metabolites selected for screening from the panel of 
drugs and drug metabolites to be tested under the FFD program and 
alcohol and its metabolites.
    (2) The instrument must be operated in accordance with the 
manufacturer's specifications. If screening detects the presence of 
drugs, drug metabolites, or alcohol at or above the instrument set 
point(s), the individual screened by the instrument must be subject to 
a POCTA screening test using the process described in paragraph (h) of 
this section or an oral fluid or urine test that is sent to an HHS-
certified laboratory.
    (3) A part 26 sanction may not be issued to an individual based 
solely on a portal area screening instrument detection that drugs or 
alcohol exceed the instrument's established setpoint.
    (k) Blood testing. The testing of blood specimens may only be 
conducted under the order of the licensee- or other entity-designated 
MRO for a valid medical reason as confirmed by the MRO pursuant to 
Sec.  26.31(d)(5). This specimen must be subject to testing by a 
laboratory that satisfies quality control requirements that are 
comparable to those required for certification by the HHS.
    (l) Custody-and-control form. For the collection and packaging of 
urine, oral fluid, and hair specimens, the licensee or other entity 
must use a custody-and-control form approved by the U.S. Office of 
Management and Budget. For the use of a POCTA device, the licensee or 
other entity must implement a licensee- or other entity-approved and -
maintained procedure that ensures the reliability of the tracking, 
handling, and storage of a specimen from the point of specimen 
collection to the final disposition of the specimen and the reliability 
of an identification system to uniquely assign the specimen to the 
donor.
    (m) Medical Review Officer. Licensees or other entities must--
    (1) Require their designated MRO to review positive, adulterated, 
substituted, and dilute confirmatory drug and validity test results and 
test results of questionable validity to determine whether the donor 
has violated the FFD policy for urine and oral fluid specimens. The 
review must be completed before reporting the results to the individual 
designated by the licensee or other entity to assess authorization or 
perform the suitability and fitness determinations required under Sec.  
26.619, or, if required, that are described in subpart H of this part.
    (2) Require their MRO to satisfy the requirements in Sec.  26.183 
and, prior to conducting any activities under this part, attend and 
pass a medical- or clinical-based training session to improve his/her 
knowledge of MRO duties and responsibilities, drug and alcohol testing 
processes and procedures, and evaluation of drug testing results. This 
training session must be conducted by a nationally recognized MRO 
training and certification organization that has been assessed by the 
licensee's or other entity's FFD program personnel to include the 
technical elements an MRO must implement under Sec.  26.185. An

[[Page 87036]]

MRO who performed the duties and responsibilities in Sec. Sec.  26.185 
and 26.187 for at least 3 continuous years in the last 10 years prior 
to being hired or contracted by the licensee or other entity satisfies 
the requirements in this paragraph.
    (3) Require their MRO to attend a medical- or clinical-based 
training session on a triennial basis to improve his/her knowledge of 
changes in drug and alcohol testing processes and procedures and 
evaluation of drug testing results.
    (4) Require their MRO to determine whether a biological specimen is 
positive, adulterated, substituted, dilute or of questionable validity 
by implementing the requirements in Sec.  26.185 or the HHS Guidelines 
through the licensee's or other entity's procedures.
    (i) If Sec.  26.185 or the HHS Guidelines, as used by the licensee 
or other entity in its procedures, are insufficient to make this 
determination, then guidance issued by a State agency in the state in 
which the NRC-licensed facility is located, Federal agencies, or 
nationally recognized MRO training and certification organizations may 
be used to inform an MRO determination.
    (ii) An MRO need not review a confirmed alcohol positive test 
result determined by an evidentiary breath testing device under 
paragraphs (c)(3)(vi) and (vii) of this section.
    (5) Require their MRO to determine and approve the use of oral 
fluid or urine as an alternative biological specimen when the donor 
cannot provide a specimen for testing. This determination and the 
retest must be documented and completed as soon as reasonably 
practicable.
    (6) Require the MRO to review all specimens screened and tested 
associated with a drug-related FFD policy violation. This review 
includes POCTA, split specimens, and all specimens taken to resolve a 
discrepant condition, such as a possible subversion attempt, impairment 
without a known cause, or a donor-requested or MRO-directed re-test. To 
resolve a discrepant condition, the MRO is authorized to test a 
specimen for a biological marker, adulterants, or additional drugs.
    (n) Limitations of screening and testing. Specimens collected under 
NRC regulations may only be designated or approved for screening and 
testing as described in this part and may not be used to conduct any 
other analysis or test without the written permission of the donor. 
Analyses, screens, and tests that may not be conducted include, but are 
not limited to, DNA testing, serological typing, or any other medical 
or genetic test used for diagnostic or specimen identification 
purposes. No biological specimens may be passively sampled and analyzed 
in a manner different than described in this subpart.
    (o) Specimen collectors. All onsite specimen collections, except a 
collection by a portal area screening instrument in paragraph (j) of 
this section, must be conducted by licensee- or other entity-designated 
and -trained personnel.


Sec.  26.608  FFD program training.

    (a) FFD program training. (1) Individuals must be trained in the 
FFD policy and procedure, including fatigue management, and their FFD 
program responsibilities. Individuals who collect specimens for testing 
or screening must also be trained in specimen collector duties and 
responsibilities, including, at a minimum, specimen collection, custody 
and control, identification and response to subversion attempts, and 
privacy. For licensees and other entities of commercial nuclear plants, 
the FFD program training program must use a systems approach to 
training as defined in Sec.  53.725 of this chapter and described in 
Sec.  53.830 of this chapter for those individuals in Sec.  26.4.
    (2) FFD program training must include training on the behavioral 
observation program. The behavioral observation program training must 
include the detection of physiological behaviors or conditions that may 
indicate--
    (i) Possible use, sale, or possession of illegal drugs or illicit 
drugs, or substance abuse on- or offsite;
    (ii) Use or possession of alcohol onsite or use while on duty 
offsite;
    (iii) Impairment from fatigue or any cause that, if left 
unattended, could result in inattentiveness or human errors; and
    (iv) Any individual's inability to safely and competently perform 
assigned duties and responsibilities or act in a trustworthy and 
reliable manner while having access to protected areas, SNM, or 
sensitive information.
    (3) Training must explain that an individual's FFD policy violation 
will--
    (i) Subject the individual to an FFD program-required sanction 
designed to preclude recurrence of an FFD policy violation;
    (ii) Contribute to the licensee's or other entity's assessment of 
whether the individual can be trusted and relied upon to safely and 
competently perform the assigned duties and responsibilities making the 
individual subject to this subpart;
    (iii) Be used to inform the licensee's or other entity's insider 
mitigation and access authorization programs under Sec. Sec.  73.55, 
73.56, 73.100 or 73.120 of this chapter; and
    (iv) Be used to inform other NRC licensees and other entities 
subject to part 26 when FFD program information is requested to support 
authorization determinations under subpart C of this part or Sec. Sec.  
73.56 or 73.120 of this chapter.
    (b) Training and assessments. Training and a trainee assessment 
must be conducted before pre-access testing, and refresher training and 
trainee assessments must be conducted periodically thereafter.
    (c) Training program review. The licensee or other entity must 
periodically evaluate its FFD training program and revise it as 
appropriate to reflect industry experience as well as applicable 
changes to the regulations in this part, the HHS Guidelines, if used, 
and specimen collection and testing processes implemented by the 
licensee or other entity.


Sec.  26.609  Behavioral observation.

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

[[Page 87037]]

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


Sec.  26.610  Sanctions.

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


Sec.  26.611  Protection of information.

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


Sec.  26.613  Appeals process.

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


Sec.  26.615  Audits.

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


Sec.  26.617  Recordkeeping and reporting.

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


Sec.  26.619  Suitability and fitness determinations.

    Licensees and other entities that implement FFD programs under this 
subpart must develop, implement, and maintain procedures for evaluating 
whether to assign individuals to perform or direct those duties and 
responsibilities making them subject to this subpart. A suitability or 
fitness determination conducted for cause must be performed face to 
face. A suitability or fitness determination conducted for cause may be 
performed remotely using electronic communications only when supported 
by someone who is present in-person with the individual being assessed, 
and that supporting person must be trained in accordance with the 
requirements of either Sec. Sec.  26.29 or 26.608.
0
94. Revise Sec.  26.709 to read as follows:

[[Page 87038]]

Sec.  26.709  Applicability.

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


Sec.  26.711  [Amended]

0
95. In Sec.  26.711, in paragraphs (c) and (d), remove the phrase ``(c) 
and (d),'' and add in its place the phrase ``(c), (d), and (f),''.


Sec.  26.825  [Amended]

0
96. In Sec.  26.825, in paragraph (b) add remove the phrase 
``Sec. Sec.  26.1, 26.3, 26.5, 26.7, 26.8, 26.9, 26.11, 26.51, 26.81, 
26.121, 26.151, 26.181, 26.201, 26.823, and 26.825'' and add in its 
place the phrase ``Sec. Sec.  26.1, 26.3, 26.5, 26.7, 26.8, 26.9, 
26.11, 26.51, 26.81, 26.121, 26.151, 26.181, 26.201, 26.601, 26.823, 
and 26.825''.

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

0
97. The authority citation for part 30 continues to read as follows:

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

0
98. In Sec.  30.4, revise the definition for ``Utilization facility'' 
to read as follows:


Sec.  30.4  Definitions.

* * * * *
    Utilization facility means a utilization facility as defined in the 
regulations contained in part 50 or part 53 of this chapter;
0
99. In Sec.  30.50, revise paragraph (c)(3) to read as follows:


Sec.  30.50  Reporting requirements.

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

PART 40--DOMESTIC LICENSING OF SOURCE MATERIAL

0
100. The authority citation for part 40 continues to read as follows:

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

0
101. In Sec.  40.60, revise paragraph (c)(3) to read as follows:


Sec.  40.60  Reporting requirements.

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

PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION 
FACILITIES

0
102. The authority citation for part 50 continues to read as follows:

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

0
103. In Sec.  50.47, revise paragraphs (a)(1) and (e) to read as 
follows:


Sec.  50.47  Emergency plans.

    (a)(1)(i) Except as provided in paragraph (d) of this section, no 
initial operating license for a nuclear power reactor will be issued 
under this part or under part 53 of this chapter unless a finding is 
made by the NRC that there is reasonable assurance that adequate 
protective measures can and will be taken in the event of a 
radiological emergency. No finding under this section is necessary for 
issuance of a renewed nuclear power reactor operating license.
    (ii) No initial combined license under parts 52 or 53 of this 
chapter will be issued unless a finding is made by the NRC that there 
is reasonable assurance that adequate protective measures can and will 
be taken in the event of a radiological emergency. No finding under 
this section is necessary for issuance of a renewed combined license.
    (iii) If an application for an early site permit under subpart A of 
part 52 of this chapter includes complete and integrated emergency 
plans under Sec.  52.17(b)(2)(ii) of this chapter or an application for 
an early site permit under subpart H of part 53 of this chapter 
includes complete and integrated emergency plans under Sec.  
53.1146(b)(2)(ii) of this chapter, no early site permit will be issued 
unless a finding is made by the NRC that the emergency plans provide 
reasonable assurance that adequate protective measures can and will be 
taken in the event of a radiological emergency.
    (iv) If an application for an early site permit proposes major 
features of the emergency plans under Sec. Sec.  52.17(b)(2)(i) or 
53.1146(b)(2)(i) of this chapter, no early site permit will be issued 
unless a finding is made by the NRC that the major features are 
acceptable in accordance with the applicable standards of either Sec.  
50.47 and appendix E to this part or the applicable requirements of 
Sec.  50.160, within the scope of emergency preparedness matters 
addressed in the major features.
* * * * *
    (e) Notwithstanding the requirements of paragraph (b) of this 
section and the provisions of Sec.  52.103 or Sec.  53.1452 of this 
chapter, a holder of a combined license under part 52 or part 53 of 
this chapter, as applicable, that is complying with the requirements of 
Sec.  50.47(b) and appendix E to this part may not load fuel or operate 
except as provided in accordance with appendix E to this part and Sec.  
50.54(gg), and a holder of a combined license under part 52 or part 53 
of this chapter that is complying with the requirements of Sec.  50.160 
may not load fuel or operate except as provided in accordance with 
Sec.  50.160(c)(2) and Sec.  50.54(gg).
* * * * *
0
104. In Sec.  50.54, revise paragraphs (q)(2), (q)(4), and (gg)(1) 
introductory text to read as follows:


Sec.  50.54  Conditions of licenses.

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

[[Page 87039]]

of this chapter, as applicable, shall follow and maintain the 
effectiveness of an emergency plan that meets the requirements in 
appendix E to this part and, for nuclear power reactor licensees, the 
planning standards of Sec.  50.47(b).
    (ii) A holder of a license under this part for a non-power 
production or utilization facility, a holder of a license under this 
part or part 53 of this chapter for a small modular reactor or a non-
light-water reactor, or a holder of a combined license under parts 52 
or 53 of this chapter after the Commission makes the finding under 
Sec. Sec.  52.103(g) or 53.1452(g) of this chapter, as applicable, for 
a small modular reactor or a non-light-water reactor, shall follow and 
maintain the effectiveness of either an emergency plan that meets the 
requirements in Sec.  50.160 or an emergency plan that meets the 
requirements in appendix E to this part and, for nuclear power reactor 
licensees, the planning standards of Sec.  50.47(b).
* * * * *
    (4) The changes to a licensee's emergency plan that reduce the 
effectiveness of the plan as defined in paragraph (q)(1)(iv) of this 
section may not be implemented without prior approval by the NRC. A 
licensee desiring to make such a change shall submit an application for 
an amendment to its license. In addition to the filing requirements of 
Sec. Sec.  50.90 and 50.91 or Sec. Sec.  53.1510 and 53.1515 of this 
chapter, as applicable, the request must include all emergency plan 
pages affected by that change and must be accompanied by a forwarding 
letter identifying the change, the reason for the change, and the basis 
for concluding that the licensee's emergency plan, as revised, will 
continue to meet either the requirements in Sec.  50.160 or the 
requirements in appendix E to this part and, for nuclear power reactor 
licensees, the planning standards of Sec.  50.47(b).
* * * * *
    (gg)(1) Notwithstanding Sec. Sec.  52.103 or 53.1452 of this 
chapter, if following the conduct of the exercise required by paragraph 
IV.f.2.a of appendix E to this part or Sec.  50.160(c)(2), as 
applicable, FEMA identifies one or more deficiencies in the state of 
offsite emergency preparedness, the holder of a combined license under 
10 CFR part 52 or under 10 CFR part 53, as applicable, may operate at 
up to 5 percent of rated thermal power only if the Commission finds 
that the state of onsite emergency preparedness provides reasonable 
assurance that adequate protective measures can and will be taken in 
the event of a radiological emergency. The NRC will base this finding 
on its assessment of the applicant's onsite emergency plans against the 
pertinent standards in either Sec.  50.47 and appendix E to this part, 
or Sec.  50.160, as applicable. Review of the applicant's emergency 
plans will include the following standards with offsite aspects:
* * * * *
0
105. In Sec.  50.160, revise paragraphs (b)(3) and (c)(2) to read as 
follows:


Sec.  50.160  Emergency preparedness for small modular reactors, non-
light-water reactors, and non-power production or utilization 
facilities.

* * * * *
    (b) * * *
    (3) Emergency planning zone. For an applicant whose analysis 
required by Sec.  50.33(g)(2) or Sec.  53.1109(g)(2) of this chapter 
meets the criteria in Sec.  50.33(g)(2)(i) or Sec.  53.1109(g)(2)(i) of 
this chapter, as applicable, determine and describe the boundary and 
physical characteristics of the EPZ in the emergency plan.
* * * * *
    (c) * * *
    (2) A holder of a combined license issued under parts 52 or 53 of 
this chapter before the Commission has made the finding under 
Sec. Sec.  52.103(g) or 53.1452(g) of this chapter, as applicable, must 
establish, implement, and maintain an emergency preparedness program 
that meets the requirements of paragraph (b) of this section, as 
described in the approved emergency plan and license, and conduct an 
initial exercise to demonstrate this compliance within 2 years before 
the scheduled date for initial loading of fuel (or, for a fueled 
manufactured reactor, within 2 years before the scheduled date for 
initiating the physical removal of any one of the independent physical 
mechanisms to prevent criticality required under Sec.  53.620(d)(1) of 
this chapter).
0
106. In appendix B to part 50, revise the first paragraph in the 
Introduction section, the first paragraph of section III, Design 
Control, and section IV, Procurement Document Control, to read as 
follows:

Appendix B to Part 50--Quality Assurance Criteria for Nuclear Power 
Plants and Fuel Reprocessing Plants

    Introduction. Every applicant for a construction permit is 
required by the provisions of Sec.  50.34 or Sec.  53.1309 of this 
chapter to include in its preliminary safety analysis report a 
description of the quality assurance program to be applied to the 
design, fabrication, construction, and testing of the structures, 
systems, and components of the facility. Every applicant for an 
operating license is required by the provisions of Sec.  50.34 or 
Sec.  53.1369 of this chapter to include, in its final safety 
analysis report, information pertaining to the managerial and 
administrative controls to be used to assure safe operation. Every 
applicant for a combined license is required by the provisions of 
Sec. Sec.  52.79 or 53.1416 of this chapter to include in its final 
safety analysis report a description of the quality assurance 
applied to the design, and to be applied to the fabrication, 
construction, and testing of the structures, systems, and components 
of the facility and to the managerial and administrative controls to 
be used to assure safe operation. For applications submitted after 
September 27, 2007, every applicant for an early site permit is 
required by the provisions of Sec. Sec.  52.17 or 53.1146 of this 
chapter to include in its site safety analysis report a description 
of the quality assurance program applied to site activities related 
to the design, fabrication, construction, and testing of the 
structures, systems, and components of a facility or facilities that 
may be constructed on the site. Every applicant for a design 
approval is required by the provisions of Sec. Sec.  52.137 or 
53.1209 of this chapter to include in its final safety analysis 
report a description of the quality assurance program applied to the 
design of the structures, systems, and components of the facility. 
Every applicant for a design certification is required by the 
provisions of Sec. Sec.  52.47 or 53.1239 of this chapter to include 
in its final safety analysis report a description of the quality 
assurance program applied to the design of the structures, systems, 
and components of the facility. Every applicant for a manufacturing 
license is required by the provisions of Sec. Sec.  52.157 or 
53.1279 of this chapter to include in its final safety analysis 
report a description of the quality assurance program applied to the 
design, and to be applied to the manufacture of, the structures, 
systems, and components of the reactor. Nuclear power plants and 
fuel reprocessing plants include structures, systems, and components 
that prevent or mitigate the consequences of postulated accidents 
that could cause undue risk to the health and safety of the public. 
This appendix establishes quality assurance requirements for the 
design, manufacture, construction, and operation of those 
structures, systems, and components. The pertinent requirements of 
this appendix apply to all activities affecting the safety-related 
functions of those structures, systems, and components; these 
activities include designing, purchasing, fabricating, handling, 
shipping, storing, cleaning, erecting, installing, inspecting, 
testing, operating, maintaining, repairing, refueling, and 
modifying.
* * * * *

III. Design Control

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

[[Page 87040]]

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

IV. Procurement Document Control

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

PART 51--ENVIRONMENTAL PROTECTION REGULATIONS FOR DOMESTIC 
LICENSING AND RELATED REGULATORY FUNCTIONS

0
107. The authority citation for part 51 continues to read as follows:

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

0
108. In Sec.  51.20, revise paragraphs (b)(1) and (2) to read as 
follows:


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

* * * * *
    (b) * * *
    (1) Issuance of a limited work authorization or a permit to 
construct a nuclear power reactor, testing facility, or fuel 
reprocessing plant under part 50 of this chapter, issuance of an early 
site permit under part 52 of this chapter, or issuance of a limited 
work authorization, construction permit, or early site permit under 
part 53 of this chapter.
    (2) Issuance or renewal of a full power or design capacity license 
to operate a nuclear power reactor, testing facility, or fuel 
reprocessing plant under parts 50 or 53 of this chapter, or a combined 
license under parts 52 or 53 of this chapter.
* * * * *
0
109. In Sec.  51.22, revise paragraphs (c)(3) introductory text, (c)(9) 
introductory text, (c)(12) introductory text, (c)(17), (c)(22) and (23) 
to read as follows:


Sec.  51.22  Criterion for categorical exclusion; identification of 
licensing and regulatory actions eligible for categorical exclusion or 
otherwise not requiring environmental review.

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


Sec.  51.26  [Amended]

0
110. In Sec.  51.26, in paragraph (d) remove the phrase ``under part 
52'' and add in its place the phrase ``under 10 CFR parts 52 or 53,''.
0
111. In Sec.  51.30, revise paragraph (a) introductory text and 
paragraphs (d) and (e) to read as follows:


Sec.  51.30  Environmental assessment.

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

[[Page 87041]]

associated with manufacturing the reactor under the manufacturing 
license.


Sec.  51.31  [Amended]

0
112. In Sec.  51.31, in paragraph (a) remove the phrase ``under part 
52'' and add in its place the phrase ``under parts 52 or 53''.


Sec.  51.32  [Amended]

0
113. In Sec.  51.32, in paragraphs (b)(1) and (3) remove the phrase 
``of part 52 of this chapter'' and add in its place the phrase ``of 
part 52 of this chapter or subpart H of part 53 of this chapter''.


Sec.  51.49  [Amended]

0
114. In Sec.  51.49, in paragraph (c) introductory text, remove the 
phrase ``of part 52 of this chapter'' and add in its place the phrase 
``of part 52 of this chapter or under subpart H of part 53 of this 
chapter''.


Sec.  51.50  [Amended]

0
115. In Sec.  51.50, wherever it appears, remove the phrase ``in 
accordance with Sec.  50.36b of this chapter'' and add in its place the 
phrase ``in accordance with Sec. Sec.  50.36b or 53.1112 of this 
chapter''.


Sec.  51.53  [Amended]

0
116. In Sec.  51.53, in paragraph (d) remove the phrase ``under Sec.  
50.82 of this chapter'' and add in its place the phrase ``under 
Sec. Sec.  50.82 or 53.1080 of this chapter''.


Sec.  51.54  [Amended]

0
117. In Sec.  51.54, in paragraph (a), remove the phrase ``of part 52 
of this chapter'' and add in its place the phrase ``of part 52 of this 
chapter or under subpart H of part 53 of this chapter''.


Sec.  51.55  [Amended]

0
118. In Sec.  51.55, in paragraph (a) remove the phrase ``of part 52 of 
this chapter'' and add in its place the phrase ``of part 52 of this 
chapter or under subpart H of part 53 of this chapter''.
0
119. In Sec.  51.58, revise paragraph (b) to read as follows:


Sec.  51.58  Environmental report--number of copies; distribution.

* * * * *
    (b) Each applicant for a license to manufacture a nuclear power 
reactor, or for an amendment to a license to manufacture, seeking 
approval of the final design of the nuclear power reactor under subpart 
F of part 52 of this chapter or under subpart H of part 53 of this 
chapter, shall submit to the Commission an environmental report or any 
supplement to an environmental report in the manner specified in 
Sec. Sec.  52.3 or 53.040 of this chapter. The applicant shall maintain 
the capability to generate additional copies of the environmental 
report or any supplement to the environmental report for subsequent 
distribution to parties and Boards in the NRC proceeding; Federal, 
State, and local officials; and any affected Indian Tribes, in 
accordance with written instructions issued by the Director, Office of 
Nuclear Reactor Regulation.
0
120. In Sec.  51.77, revise paragraph (a) introductory text to read as 
follows:


Sec.  51.77  Distribution of draft environmental impact statement.

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


Sec.  51.92  [Amended]

0
121. In Sec.  51.92, in paragraph (b), wherever it may appear, remove 
the phrase ``10 CFR part 52'' and add in its place the phrase ``10 CFR 
parts 52 or 53''.


Sec.  51.95  [Amended]

0
122. In Sec.  51.95, in paragraph (c) introductory text remove the 
phrase ``under 10 CFR parts 52 or 54'' and add in its place the phrase 
``under 10 CFR parts 52, 53, or 54''.
0
123. In Sec.  51.101, revise paragraph (a)(2) to read as follows:


Sec.  51.101  Limitations on actions.

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


Sec.  51.103  [Amended]

0
124. In Sec.  51.103, in paragraph (a)(6) remove the phrase ``under 10 
CFR 50.10'' and add in its place the phrase ``under Sec. Sec.  50.10 or 
53.1130 of this chapter''.
0
125. In Sec.  51.105, revise paragraph (c)(1) introductory text to read 
as follows:


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

* * * * *
    (c)(1) In addition to complying with the applicable provisions of 
Sec.  51.104, in any proceeding for the issuance of a construction 
permit for a nuclear power plant or an early site permit under parts 52 
or 53 of this chapter, where the applicant requests a limited work 
authorization under Sec. Sec.  50.10(d) or 53.1130 of this chapter, the 
presiding officer will--
* * * * *
0
126. In Sec.  51.107, revise paragraphs (a) introductory text, (b) 
introductory text, and (d)(1) introductory text to read as follows:


Sec.  51.107  Public hearings in proceedings for issuance of combined 
licenses; limited work authorizations.

    (a) In addition to complying with the applicable requirements of 
Sec.  51.104, in a proceeding for the issuance of a combined license 
for a nuclear power reactor under parts 52 or 53 of this chapter, the 
presiding officer will:
* * * * *
    (b) If a combined license application references an early site 
permit, then the presiding officer in the combined license hearing must 
not admit any contention proffered by any party on environmental issues 
that have been accorded finality under Sec. Sec.  52.39 or 53.1188 of 
this chapter, unless the contention:
* * * * *
    (d)(1) In any proceeding for the issuance of a combined license 
where the applicant requests a limited work authorization under 
Sec. Sec.  50.10(d) or Sec.  53.1130(a) of this chapter, the presiding 
officer, in addition to complying with any applicable provision of 
Sec.  51.104, will:
* * * * *
0
127. Revise Sec.  51.108 to read as follows:


Sec.  51.108  Public hearings on Commission findings that inspections, 
tests, analyses, and acceptance criteria of combined licenses are met.

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

[[Page 87042]]

subpart H of part 53 of this chapter, or the adequacy of any other 
environmental impact statement or environmental assessment referenced 
in the combined license application. The Commission will not make any 
environmental findings in connection with the finding under Sec.  
52.103(g) or Sec.  53.1452(g) of this chapter.
0
128. Add part 53, consisting of Sec. Sec.  53.000 through 53.9010, to 
read as follows:

PART 53--RISK-INFORMED, TECHNOLOGY-INCLUSIVE REGULATORY FRAMEWORK 
FOR COMMERCIAL NUCLEAR PLANTS

Sec.
53.000 Purpose.
Subpart A--General Provisions
53.015 Scope.
53.020 Definitions.
53.030 Reserved.
53.040 Written communications.
53.050 Deliberate misconduct.
53.060 Employee protection.
53.070 Completeness and accuracy of information.
53.080 Specific exemptions.
53.090 Standards for review.
53.100 Jurisdictional limits.
53.110 Attacks and destructive acts.
53.115 Rights related to special nuclear material.
53.117 License suspension and rights of recapture.
53.120 Information collection requirements: OMB approval.
Subpart B--Technology-Inclusive Safety Requirements
53.210 Safety criteria for design-basis accidents.
53.220 Safety criteria for licensing-basis events other than design-
basis accidents.
53.230 Safety functions.
53.240 Licensing-basis events.
53.250 Defense in depth.
53.260 Normal operations.
53.270 Protection of plant workers.
Subpart C--Design and Analysis Requirements
53.400 Design features for licensing-basis events.
53.410 Functional design criteria for design-basis accidents.
53.415 Protection against external hazards.
53.420 Functional design criteria for licensing-basis events other 
than design-basis accidents.
53.425 Design features and functional design criteria for normal 
operations.
53.430 Design features and functional design criteria for protection 
of plant workers.
53.440 Design requirements.
53.450 Analysis requirements.
53.460 Safety categorization and treatments.
53.470 Maintaining analytical safety margins used to justify 
operational flexibilities.
53.480 Earthquake engineering.
Subpart D--Siting Requirements
53.500 General siting and siting assessment.
53.510 External hazards.
53.520 Site characteristics.
53.530 Population-related considerations.
53.540 Siting interfaces.
Subpart E--Construction and Manufacturing Requirements
53.600 Construction and manufacturing--scope and purpose.
53.605 Reporting of defects and noncompliance.
53.610 Construction.
53.620 Manufacturing.
Subpart F--Requirements for Operation
53.700 Operational objectives.
53.710 Maintaining capabilities and availability of structures, 
systems, and components.
53.715 Maintenance, repair, and inspection programs.
53.720 Response to seismic events.
53.725 General staffing, training, personnel qualifications, and 
human factors requirements.
53.726 Communications.
53.728 Completeness and accuracy of information.
53.730 Defining, fulfilling, and maintaining the role of personnel 
in ensuring safe operations.
53.735 General exemptions.
53.740 Facility licensee requirements--General.
53.745 Operator license requirements.
53.760 Operator licensing.
53.765 Medical requirements.
53.770 Incapacitation because of disability or illness.
53.775 Applications for operators and senior operators.
53.780 Training, examination, and proficiency program.
53.785 Conditions of operator and senior operator licenses.
53.790 Issuance, modification, and revocation of operator and senior 
operator licenses.
53.795 Expiration and renewal of operator and senior operator 
licenses.
53.800 Facility licensees for self-reliant-mitigation facilities.
53.805 Facility licensee requirements related to generally licensed 
reactor operators.
53.810 Generally licensed reactor operators.
53.815 Generally licensed reactor operator training, examination, 
and proficiency programs.
53.820 Cessation of individual applicability.
53.830 Training and qualification of commercial nuclear plant 
personnel.
53.845 Programs.
53.850 Radiation protection.
53.855 Emergency preparedness.
53.860 Security programs.
53.865 Quality assurance.
53.870 Integrity assessment programs.
53.875 Fire protection.
53.880 Inservice inspection and inservice testing.
53.910 Procedures and guidelines.
Subpart G--Decommissioning Requirements
53.1000 Scope and purpose.
53.1010 Financial assurance for decommissioning.
53.1020 Cost estimates for decommissioning.
53.1030 Annual adjustments to cost estimates for decommissioning.
53.1040 Methods for providing financial assurance for 
decommissioning.
53.1045 Limitations on the use of decommissioning trust funds.
53.1050 NRC oversight.
53.1060 Reporting and recordkeeping requirements.
53.1070 Termination of license.
53.1075 Program requirements during decommissioning.
53.1080 Release of part of a commercial nuclear plant or site for 
unrestricted use.
Subpart H--Licenses, Certifications, and Approvals
53.1100 Filing of application for licenses, certifications, or 
approvals; oath or affirmation.
53.1101 Requirement for license.
53.1103 Combining applications and licenses.
53.1106 Elimination of repetition.
53.1109 Contents of applications; general information.
53.1112 Environmental conditions.
53.1115 Agreement limiting access to classified information.
53.1118 Ineligibility of certain applicants.
53.1120 Exceptions and exemptions from licensing requirements.
53.1121 Public inspection of applications.
53.1124 Relationship between sections.
53.1130 Limited work authorizations.
53.1140 Early site permits.
53.1143 Filing of applications.
53.1144 Contents of applications for early site permits; general 
information.
53.1146 Contents of applications for early site permits; technical 
information.
53.1149 Review of applications.
53.1155 Referral to the Advisory Committee on Reactor Safeguards.
53.1158 Issuance of early site permit.
53.1161 Extent of activities permitted.
53.1164 Duration of permit.
53.1167 Limited work authorization after issuance of early site 
permit.
53.1170 Transfer of early site permit.
53.1173 Application for renewal.
53.1176 Criteria for renewal.
53.1179 Duration of renewal.
53.1182 Use of site for other purposes.
53.1188 Finality of early site permit determinations.
53.1200 Standard design approvals.
53.1203 Filing of applications.
53.1206 Contents of applications for standard design approvals; 
general information.
53.1209 Contents of applications for standard design approvals; 
technical information.
53.1210 Contents of applications for standard design approvals; 
other application content.
53.1212 Standards for review of applications.

[[Page 87043]]

53.1215 Referral to the Advisory Committee on Reactor Safeguards.
53.1218 Staff approval of design.
53.1221 Finality of standard design approvals; information requests.
53.1230 Standard design certifications.
53.1233 Filing of applications.
53.1236 Contents of applications for standard design certifications; 
general information.
53.1239 Contents of applications for standard design certifications; 
technical information.
53.1241 Contents of applications for standard design certifications; 
other application content.
53.1242 Review of applications.
53.1245 Referral to the Advisory Committee on Reactor Safeguards.
53.1248 Issuance of standard design certification.
53.1251 Duration of certification.
53.1254 Application for renewal.
53.1257 Criteria for renewal.
53.1260 Duration of renewal.
53.1263 Finality of standard design certifications.
53.1270 Manufacturing licenses.
53.1273 Filing of applications.
53.1276 Contents of applications for manufacturing licenses; general 
information.
53.1279 Contents of applications for manufacturing licenses; 
technical information.
53.1282 Contents of applications for manufacturing licenses; other 
application content.
53.1285 Review of applications.
53.1286 Referral to the Advisory Committee on Reactor Safeguards.
53.1287 Issuance of manufacturing licenses.
53.1288 Finality of manufacturing licenses.
53.1291 Duration of manufacturing licenses.
53.1293 Transfer of manufacturing licenses.
53.1295 Renewal of manufacturing licenses.
53.1300 Construction permits.
53.1306 Contents of applications for construction permits; general 
information.
53.1309 Contents of applications for construction permits; technical 
information.
53.1312 Contents of applications for construction permits; other 
application content.
53.1315 Review of applications.
53.1318 Finality of referenced NRC approvals, permits, and 
certifications.
53.1324 Referral to the Advisory Committee on Reactor Safeguards.
53.1327 Authorization to conduct limited work authorization 
activities.
53.1330 Exemptions, departures, and variances.
53.1333 Issuance of construction permits.
53.1336 Finality of construction permits.
53.1342 Duration of construction permits.
53.1345 Transfer of construction permits.
53.1348 Termination of construction permits.
53.1360 Operating licenses.
53.1366 Contents of applications for operating licenses; general 
information.
53.1369 Contents of applications for operating licenses; technical 
information.
53.1372 Contents of applications for operating licenses; other 
application content.
53.1375 Review of applications.
53.1381 Referral to the Advisory Committee on Reactor Safeguards.
53.1384 Exemptions, departures, and variances.
53.1387 Issuance of operating licenses.
53.1390 Backfitting of operating licenses.
53.1396 Duration of operating licenses.
53.1399 Transfer of an operating license.
53.1402 Application for renewal.
53.1405 Continuation of an operating license.
53.1410 Combined licenses.
53.1413 Contents of applications for combined licenses; general 
information.
53.1416 Contents of applications for combined licenses; technical 
information.
53.1419 Contents of applications for combined licenses; other 
application content.
53.1422 Review of applications.
53.1425 Finality of referenced NRC approvals.
53.1431 Referral to the Advisory Committee on Reactor Safeguards.
53.1434 Authorization to conduct limited work authorization 
activities.
53.1437 Exemptions, departures, and variances.
53.1440 Issuance of combined licenses.
53.1443 Finality of combined licenses.
53.1449 Inspection during construction.
53.1452 Operation under a combined license.
53.1455 Duration of combined license.
53.1456 Transfer of a combined license.
53.1458 Application for renewal.
53.1461 Continuation of combined license.
53.1470 Standardization of commercial nuclear plant designs: 
licenses to construct and operate nuclear power reactors of 
identical design at multiple sites.
Subpart I--Maintaining and Revising Licensing-Basis Information
53.1500 Licensing-basis information.
53.1502 Specific terms and conditions of licenses.
53.1505 Changes to licensing-basis information requiring prior NRC 
approval.
53.1510 Application for amendment of license.
53.1515 Public notices; State consultation.
53.1520 Issuance of amendment.
53.1525 Revising certification information within a design 
certification rule.
53.1530 Revising design information within a manufacturing license.
53.1535 Amendments during construction.
53.1540 Updating licensing-basis information and determining the 
need for NRC approval.
53.1545 Updating Final Safety Analysis Reports.
53.1550 Evaluating changes to facility as described in Final Safety 
Analysis Reports.
53.1560 Updating program documents included in licensing-basis 
information.
53.1565 Evaluating changes to programs included in licensing-basis 
information.
53.1570 Transfer of licenses.
53.1575 Termination of licenses.
53.1580 Information requests.
53.1585 Revocation, suspension, modification of licenses and 
approvals for cause.
53.1590 Backfitting.
53.1595 Renewal.
Subpart J--Reporting and Other Administrative Requirements
53.1600 General information.
53.1610 Unfettered access for inspections.
53.1620 Maintenance of records, making of reports.
53.1630 Immediate notification requirements for operating commercial 
nuclear plants.
53.1640 Licensee event report system.
53.1645 Reports of radiation exposure to members of the public.
53.1650 Facility information and verification.
53.1660 Financial requirements.
53.1670 Financial qualifications.
53.1680 Annual financial reports.
53.1690 Licensee's change of status; financial qualifications.
53.1700 Creditor regulations.
53.1710 Financial protection.
53.1720 Insurance required to stabilize and decontaminate plant 
following an accident.
53.1730 Financial protection requirements.
Subpart M--Enforcement
53.9000 Violations.
53.9010 Criminal penalties.

    Authority:  Atomic Energy Act of 1954, secs. 11, 101, 103, 108, 
122, 147, 161, 181, 182, 183, 184, 185, 186, 187, 189, 223, 234 (42 
U.S.C. 2014, 2131, 2132, 2133, 2134, 2135, 2138, 2152, 2167, 2169, 
2201, 2231, 2232, 2233, 2234, 2235, 2236, 2237, 2239, 2273, 2282); 
Energy Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 
U.S.C. 5841, 5842, 5846, 5851); Nuclear Waste Policy Act of 1982, 
sec. 306 (42 U.S.C. 10226); National Environmental Policy Act of 
1969 (42 U.S.C. 4332); 44 U.S.C. 3504 note; Sec. 109, Pub. L. 96-
295, 94 Stat. 783; Pub. L. 115-439, 132 Stat. 5571.


Sec.  53.000  Purpose.

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

[[Page 87044]]

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

Subpart A--General Provisions


Sec.  53.015  Scope.

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


Sec.  53.020  Definitions.

    For the purpose of this part:
    Anticipated event sequence means event sequences expected to occur 
one or more times during the life of a commercial nuclear plant. 
Anticipated event sequences take into account the expected response of 
all structures, systems, and components (SSCs) within the plant, 
regardless of safety classification.
    Applicant means a person applying for a license, permit, or other 
form of Commission permission or approval under this part.
    Certified fuel handler means, for a commercial nuclear plant, 
either--
    (1) A non-licensed operator who has qualified in accordance with a 
fuel handler training program approved by the Commission; or
    (2) A non-licensed operator who demonstrates compliance with the 
following criteria:
    (i) Has qualified in accordance with a fuel handler training 
program that demonstrates compliance with the same requirements as 
training programs for non-licensed operators required by Sec.  53.830, 
and
    (ii) Is responsible for decisions on--
    (A) Safe conduct of decommissioning activities,
    (B) Safe handling and storage of spent fuel, and
    (C) Appropriate response to plant emergencies.
    Combined license (COL) means a combined construction permit (CP) 
and operating license (OL) with conditions for a commercial nuclear 
plant issued under this part.
    Commercial nuclear plant means a facility consisting of one or more 
commercial nuclear reactors and associated co-located support 
facilities, including the collection of buildings, radionuclide 
sources, and SSCs for which a license, certification, or approval is 
being sought under this part, that is or will be used for producing 
power for commercial electric power or other commercial purposes. For 
the purposes of requirements in this part that reference requirements 
in part 50 of this chapter, a commercial nuclear plant is equivalent to 
a nuclear power plant.
    Commercial nuclear reactor means an apparatus, other than an atomic 
weapon, designed or used to sustain nuclear fission. For the purposes 
of requirements in this part that reference requirements in 10 CFR part 
50, a commercial nuclear reactor is equivalent to a nuclear reactor as 
defined in 10 CFR 50.2.
    Commission means the U.S. Nuclear Regulatory Commission (NRC) or 
its duly authorized representatives.
    Consensus code or standard means any technical standard that is--
    (1) Developed or adopted by a voluntary consensus standard body 
under procedures that assure that persons having interests within the 
scope of the standard that are affected by the provisions of the 
standard have reached substantial agreement on its adoption;
    (2) Formulated in a manner that afforded an opportunity for diverse 
views to be considered; and
    (3) Designated by the standards body as a consensus code or 
standard.
    Construction means the activities in paragraph (1) below and does 
not mean the activities in paragraph (2) below.
    (1) Activities constituting construction are those activities 
credited or relied upon for demonstrating compliance with the safety 
criteria defined in subpart B of this part which are conducted on-site 
to build the commercial nuclear plant, including the driving of piles; 
subsurface preparation; placement of backfill, concrete, or permanent 
retaining walls within an excavation; installation of foundations; or 
in-place assembly, erection, fabrication, or testing, which are for--
    (i) Safety-related (SR) and non-safety-related but safety-
significant (NSRSS) SSCs of a facility;
    (ii) SSCs necessary to comply with 10 CFR part 73; or
    (iii) Onsite emergency facilities necessary to comply with Sec.  
53.855.
    (2) Construction does not include--
    (i) Changes for temporary use of the land for public recreational 
purposes;
    (ii) Site exploration, including necessary borings to determine 
foundation conditions or other preconstruction monitoring to establish 
background information related to the suitability of the site, the 
environmental impacts of construction or operation, or the protection 
of environmental values;
    (iii) Preparation of a site for construction of a facility, 
including clearing of the site, grading, installation of drainage, 
erosion, and other environmental mitigation measures, and construction 
of temporary roads and borrow areas;
    (iv) Erection of fences and other access control measures;
    (v) Excavation;
    (vi) Erection of support buildings (such as construction equipment 
storage sheds, warehouse and shop facilities, utilities, concrete 
mixing plants, docking and unloading facilities, and office buildings) 
for use in connection with the construction of the facility;
    (vii) Building of service facilities (such as paved roads, parking 
lots, railroad spurs, exterior utility and lighting systems, potable 
water systems, sanitary sewage treatment facilities, and transmission 
lines);
    (viii) Procurement or fabrication of components or portions of the 
proposed facility occurring at locations other than the final, in-place 
location at the facility; or
    (ix) Manufacture of a nuclear power reactor under a manufacturing 
license (ML) under subpart H of this part to be installed at the 
proposed site and to be part of the proposed facility.
    Custom combined license (custom COL) means a COL that does not 
reference a standard design approval or design certification.
    Decommission or decommissioning means to remove a plant or site 
safely from service and reduce residual radioactivity to a level that 
permits--
    (1) Release of the property for unrestricted use and termination of 
the license; or
    (2) Release of the property under restricted conditions and 
termination of the license.
    Defense in depth means inclusion of two or more independent and 
redundant layers of defense in the design of a facility and its 
operating procedures to compensate for uncertainties such that no 
single layer of defense, no matter how robust, is exclusively relied 
upon. Defense in depth includes, but is not limited to, the use of 
access controls, physical barriers, redundant and diverse safety 
functions, and emergency response measures.
    Design-basis accidents (DBAs) means postulated event sequences that 
are used to set functional design criteria and performance objectives 
for the design of SR SSCs through deterministic analyses. Design-basis 
accidents are a type of licensing-basis event and are based on the 
capabilities and reliabilities of SR SSCs needed to mitigate and 
prevent event sequences, respectively.
    Design-basis external hazard level means the level of severity or 
intensity

[[Page 87045]]

of an external hazard for which the SR SSCs are protected against or 
designed to withstand without losing their capability to perform their 
safety functions.
    Design features means the active and passive SSCs and the inherent 
characteristics of those SSCs that contribute to limiting the total 
effective dose equivalent to individual members of the public during 
normal operations and prevent or mitigate the consequences of event 
sequences.
    Electric utility means any entity that generates or distributes 
electricity and that recovers the cost of this electricity, either 
directly or indirectly, through rates established by the entity itself 
or by a separate regulatory authority. Investor-owned utilities, 
including generation or distribution subsidiaries, public utility 
districts, municipalities, rural electric cooperatives, and State and 
Federal agencies, including associations of any of the foregoing, are 
included within the meaning of ``electric utility.''
    Event sequence means a postulated initiating event defined for a 
set of initial plant conditions followed by system, safety function, 
and operator successes or failures, and terminating in a specified end 
state depending on the system, safety function, and operator successes 
and failures (e.g., prevention of release of radioactive material or 
release in one of the reactor-specific release categories). An event 
sequence may include many unique variations of events that are similar 
in terms of results or end states.
    Exclusion area means that area surrounding the reactor, in which 
the reactor licensee has the authority to determine all activities 
including exclusion or removal of personnel and property from the area. 
This area may be traversed by a highway, railroad, or waterway, 
provided these are not so close to the facility as to interfere with 
normal operations of the facility and provided appropriate and 
effective arrangements are made to control traffic on the highway, 
railroad, or waterway, in case of emergency, to protect the public 
health and safety. Residence within the exclusion area must normally be 
prohibited. In any event, residents must be subject to ready removal in 
case of necessity. Activities unrelated to operation of the reactor may 
be permitted in an exclusion area under appropriate limitations, 
provided that no significant hazards to the public health and safety 
will result.
    Fission product release means the amount and composition of 
radioactive material released to the environment, after accounting for 
any retention of radionuclides provided by reactor design features.
    Fuel means special nuclear material (SNM) or source material, 
discrete elements that physically contain SNM or source material, and 
homogeneous mixtures that contain SNM or source material, intended to 
or used to create power in a commercial nuclear plant.
    Functional design criteria means metrics for the performance of 
SSCs. For SR SSCs, these criteria define performance metrics necessary 
to demonstrate compliance with the safety criteria in Sec.  53.210. For 
NSRSS SSCs, these criteria define performance metrics necessary to 
demonstrate compliance with the safety criteria in Sec.  53.220.
    License, when used in the context of a facility, means a limited 
work authorization, CP, OL, early site permit, COL, or ML under this 
part, or a renewed license issued by the Commission under this part. 
When used in the context of a license authorizing an individual to 
manipulate the controls of a facility, license means a license issued 
by the Commission to perform the function of an operator, senior 
operator, or generally licensed reactor operator as defined in this 
part.
    Licensee means a person who is authorized to conduct activities 
under a license issued under this part by the Commission.
    Licensing-basis events means a collection of event sequences 
considered in the design and licensing of the commercial nuclear plant. 
Licensing-basis events are unplanned events and include anticipated 
event sequences, unlikely event sequences, very unlikely event 
sequences, and DBAs.
    Licensing-basis information means the information contained in 
regulations, orders, licenses, certifications, or approvals issued by 
the NRC for a commercial nuclear plant licensed under this part and 
that information submitted to the NRC by an applicant or licensee in a 
Safety Analysis Report, program description, or other licensing-related 
document required under this part.
    Low-population zone means the area immediately surrounding the 
exclusion area which contains residents, the total number and density 
of which are such that there is a reasonable probability that 
appropriate protective measures could be taken on their behalf in the 
event of a serious accident. A permissible population density or total 
population within this zone is not included in this definition because 
the situation may vary from case to case. Whether a specific number of 
people can, for example, be evacuated from a specific area or 
instructed to take shelter on a timely basis, will depend on many 
factors such as location, number and size of highways, scope and extent 
of advance planning, and actual distribution of residents within the 
area.
    Major decommissioning activity means, for a commercial nuclear 
plant, any activity that results in permanent removal of major 
radioactive components, permanently modifies the structure of the 
containment, if applicable, or results in dismantling components for 
shipment containing greater than class C waste in accordance with 10 
CFR 61.55.
    Major feature of the emergency plans means an aspect of those plans 
necessary to:
    (1) Address in whole or part either one or more of the 16 standards 
in 10 CFR 50.47(b) or the requirements of 10 CFR 50.160(b), as 
applicable; or
    (2) Describe the emergency planning zones as required in Sec.  
53.1109(g).
    Manufactured reactor means the essential portions of a nuclear 
reactor that are manufactured under an ML and subsequently transported 
and incorporated into a commercial nuclear plant under a COL.
    Manufacturing license means a license issued under this part that 
authorizes the manufacture of manufactured reactors but not its 
construction, installation, or operation.
    Non-Safety-Related but Safety-Significant (NSRSS) SSCs means those 
SSCs which are not SR but are relied on to achieve adequate defense in 
depth or perform risk-significant functions and warrant special 
treatment.
    Non-Safety-Significant SSCs means those SSCs that are not SR or 
NSRSS, are not relied on to achieve adequate defense in depth or to 
perform risk-significant functions, and do not warrant special 
treatment.
    Person means--
    (1) any individual, corporation, partnership, firm, association, 
trust, estate, public or private institution, group, government agency 
other than the Commission or the Department, except that the Department 
shall be considered a person to the extent that its facilities are 
subject to the licensing and related regulatory authority of the 
Commission pursuant to section 202 of the ERA, any State or any 
political subdivision of, or any political entity within a State, any 
foreign government or nation or any political subdivision of any such 
government or nation, or other entity; and
    (2) any legal successor, representative, agent, or agency of the 
foregoing.

[[Page 87046]]

    Population center distance means the distance from the reactor to 
the nearest boundary of a densely populated center containing more than 
about 25,000 residents.
    Probabilistic risk assessment means a quantitative assessment of 
the risk associated with plant operation and maintenance that is 
measured in terms of event sequence occurrence frequencies and 
consequences.
    Programmatic controls means administrative procedures that govern 
human action in implementing programs and operating, monitoring, and 
maintaining SSCs and equipment of a commercial nuclear plant. 
Programmatic controls considered to be licensing basis information are 
specified in an application for a requested activity of the Commission.
    Quality assurance (QA) means all those planned and systematic 
actions necessary to ensure that a structure, system, or component will 
perform satisfactorily in service. Quality assurance includes quality 
control, which comprises those QA actions related to the physical 
characteristics of a material, structure, component, or system which 
provide a means to control the quality of the material, structure, 
component, or system to predetermined requirements.
    Safety criteria means performance-based metrics that establish a 
level of safety provided in requirements in Sec. Sec.  53.210 and 
53.220.
    Safety-related structures, systems, or components means those SSCs 
that are relied upon to demonstrate compliance with the safety criteria 
in Sec.  53.210 and warrant special treatment.
    Small modular reactor means a power reactor, which may be of 
modular design as defined in 10 CFR 52.1, licensed under this part to 
produce heat energy up to 1,000 megawatts thermal per module.
    Site characteristics means the actual physical, environmental, and 
demographic features of a site. Site characteristics are specified in 
an early site permit or in a Preliminary or Final Safety Analysis 
Report for a limited work authorization, CP, or COL, as applicable.
    Site parameters are the postulated physical, environmental, and 
demographic features of an assumed site. Site parameters are specified 
in a standard design approval, standard design certification, or ML.
    Source material means source material as defined in subsection 11z. 
of the Atomic Energy Act of 1954, as amended, (the Act) and in the 
regulations contained in part 40 of this chapter.
    Special nuclear material (SNM) means:
    (1) Plutonium, uranium-233, uranium enriched in the isotope-233 or 
in the isotope-235, and any other material which the Commission, 
pursuant to the provisions of section 51 of the Act, determines to be 
SNM, but does not include source material; or
    (2) Any material artificially enriched by any of the foregoing, but 
does not include source material.
    Special treatment means those requirements, such as QA and 
programmatic controls, that ensure that SR and NSRSS SSCs will provide 
defense in depth or perform risk-significant functions. The 
requirements also ensure that the SSCs will perform under the service 
conditions and with the reliability assumed in the analysis performed 
under Sec.  53.450 to demonstrate compliance with the safety criteria 
in Sec. Sec.  53.210 and 53.220.
    Standard design means a design which is sufficiently detailed and 
complete to support certification or approval in accordance with 
subpart H of this part, and which is usable under of this part for a 
multiple number of units or at a multiple number of sites without 
reopening or repeating the review.
    Standard design approval or design approval means an NRC staff 
approval, issued under subpart H of this part, of a final standard 
design for a commercial nuclear plant. The approval may be for either 
the final design for the entire reactor facility or the final design of 
major portions thereof.
    Standard design certification or design certification means a 
Commission approval, issued under subpart H of this part, of a final 
standard design for a nuclear power facility. This design may be 
referred to as a certified standard design.
    Total effective dose equivalent means the sum of the effective dose 
equivalent (for external exposures) and the committed effective dose 
equivalent (for internal exposures).
    Utilization facility means any commercial nuclear reactor other 
than one designed or used primarily for the formation of plutonium or 
uranium-233.
    Unlikely event sequences means event sequences that are not 
expected to occur in the life of a commercial nuclear plant and are 
less likely than anticipated event sequences, but are infrequent rather 
than rare. Unlikely event sequences take into account the expected 
response of all SSCs within the plant regardless of safety 
classification.
    Very unlikely event sequences means event sequences that are not 
expected to occur in the life of a commercial nuclear plant, are less 
likely than an unlikely event sequence, and are rare. Very unlikely 
event sequences take into account the expected response of all SSCs 
within the plant regardless of safety classification.


Sec.  53.030  [Reserved]


Sec.  53.040  Written communications.

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

[[Page 87047]]

miscellaneous requests for additional information) that are required of 
holders of licenses, permits, and design approvals issued pursuant to 
this part, must be submitted as follows, except as otherwise specified 
in paragraphs (b)(2) through (7) of this section: to the NRC's Document 
Control Desk (if on paper, the signed original), with a copy to the 
appropriate Regional Office, and a copy to the appropriate NRC Resident 
Inspector if one has been assigned to the site of the facility or the 
place of manufacture of a reactor licensed under this part.
    (2) Applications for permits and licenses, and amendments to 
applications. Applications for licenses, permits, and design approvals 
and amendments to any of these types of applications must be submitted 
to the NRC's Document Control Desk, with a copy to the appropriate 
Regional Office, and a copy to the appropriate NRC Resident Inspector 
if one has been assigned to the facility or the place of manufacture of 
a reactor licensed under this part, except as otherwise specified in 
paragraphs (b)(3) through (9) of this section. If the application or 
amendment is on paper, the submission to the Document Control Desk must 
be the signed original.
    (3) Acceptance review application. Written communications required 
for an application for determination of suitability for docketing must 
be submitted to the NRC's Document Control Desk, with a copy to the 
appropriate Regional Office. If the communication is on paper, the 
submission to the Document Control Desk must be the signed original.
    (4) Security plan and related submissions. Written communications, 
as defined in paragraphs (b)(4)(i) through (v) of this section, must be 
submitted to the NRC's Document Control Desk, with a copy to the 
appropriate Regional Office. If the communication is on paper, the 
submission to the Document Control Desk must be the signed original. 
Submissions should include the following as appropriate:
    (i) Physical security plan;
    (ii) Safeguards contingency plan;
    (iii) Cybersecurity plan;
    (iv) Change to security plan, guard training and qualification 
plan, safeguards contingency plan, or cybersecurity plan made without 
prior Commission approval under Sec.  53.1565; and
    (v) Application for amendment of physical security plan, guard 
training and qualification plan, safeguards contingency plan, or 
cybersecurity plan under Sec.  53.1510.
    (5) Emergency plan and related submissions. Written communications 
as defined in paragraphs (b)(5)(i) through (iii) of this section must 
be submitted to the NRC's Document Control Desk, with a copy to the 
appropriate Regional Office, and a copy to the appropriate NRC Resident 
Inspector if one has been assigned to the site of the facility. If the 
communication is on paper, the submission to the Document Control Desk 
must be the signed original. Submissions should include the following 
as appropriate:
    (i) Emergency plan;
    (ii) Change to an emergency plan under Sec.  53.1565; and
    (iii) Emergency implementing procedures under Sec.  53.855.
    (6) Updated Final Safety Analysis Report. An Updated Final Safety 
Analysis Report or replacement pages under Sec.  53.1545 must be 
submitted to the NRC's Document Control Desk, with a copy to the 
appropriate Regional Office, and a copy to the appropriate NRC Resident 
Inspector if one has been assigned to the site of the facility or the 
place of manufacture of a reactor licensed under this part. Paper copy 
submissions may be made using replacement pages; however, if a licensee 
chooses to use electronic submission, all subsequent updates or 
submissions must be performed electronically on a total replacement 
basis. If the communication is on paper, the submission to the Document 
Control Desk must be the signed original. If the communications are 
submitted electronically, see Guidance for Electronic Submissions to 
the Commission.
    (7) Quality assurance related submissions. (i) A change to the 
Safety Analysis Report QA program description under Sec.  53.1565, or a 
change to a licensee's NRC-accepted QA topical report under Sec.  
53.1565, must be submitted to the NRC's Document Control Desk, with a 
copy to the appropriate Regional Office, and a copy to the appropriate 
NRC Resident Inspector if one has been assigned to the site of the 
facility or the place of manufacture of a reactor licensed under this 
part. If the communication is on paper, the submission to the Document 
Control Desk must be the signed original.
    (ii) A change to an NRC-accepted QA topical report from non-
licensees (i.e., architect/engineers, nuclear steam supply system 
suppliers, fuel suppliers, constructors, etc.) must be submitted to the 
NRC's Document Control Desk. If the communication is on paper, the 
signed original must be sent.
    (8) Certification of permanent cessation of operations. The 
licensee's certification of permanent cessation of operations, under 
subpart G of this part, must state the date on which operations have 
ceased or will cease, and must be submitted to the NRC's Document 
Control Desk. This submission must be under oath or affirmation.
    (9) Certification of permanent fuel removal. The licensee's 
certification of permanent fuel removal, under subpart G of this part, 
must state the date on which the fuel was removed from the reactor 
vessel and the disposition of the fuel, and must be submitted to the 
NRC's Document Control Desk. This submission must be under oath or 
affirmation.
    (c) Form of communications. All paper copies submitted to 
demonstrate compliance with the requirements set forth in paragraph (b) 
of this section must be typewritten, printed, or otherwise reproduced 
in permanent form on unglazed paper. Exceptions to these requirements 
imposed on paper submissions may be granted for the submission of 
micrographic, photographic, or similar forms.
    (d) Regulation governing submission. Licensees, applicants, and 
holders of standard design approvals submitting correspondence, 
reports, and other written communications under the regulations of this 
part are requested but not required to cite whenever practical, in the 
upper right corner of the first page of the submission, the specific 
regulation or other basis requiring submission.


Sec.  53.050  Deliberate misconduct.

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

[[Page 87048]]

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


Sec.  53.060  Employee protection.

    (a) Discrimination by a Commission licensee, holder of a standard 
design approval, an applicant for a license, standard design 
certification, or standard design approval, a contractor or 
subcontractor of a Commission licensee, holder of a standard design 
approval, applicant for a license, standard design certification, or 
standard design approval, against an employee for engaging in certain 
protected activities is prohibited. Discrimination includes discharge 
and other actions that relate to compensation, terms, conditions, or 
privileges of employment. The protected activities are established in 
section 211 of the Energy Reorganization Act of 1974, as amended, and 
in general are related to the administration or enforcement of a 
requirement imposed under the Act or the Energy Reorganization Act of 
1974, as amended.
    (1) The protected activities include but are not limited to--
    (i) Providing the Commission or his or her employer information 
about alleged violations of either of the statutes named in paragraph 
(a) of this section or possible violations of requirements imposed 
under either of those statutes;
    (ii) Refusing to engage in any practice made unlawful under either 
of the statutes named in paragraph (a) of this section or under these 
requirements if the employee has identified the alleged illegality to 
the employer;
    (iii) Requesting the NRC to institute action against his or her 
employer for the administration or enforcement of these requirements;
    (iv) Testifying in any Commission proceeding, or before Congress, 
or at any Federal or State proceeding regarding any provision (or 
proposed provision) of either of the statutes named in paragraph (a) of 
this section; and
    (v) Assisting or participating in, or being about to assist or 
participate in, these activities.
    (2) These activities are protected even if no formal proceeding is 
actually initiated as a result of the employee assistance or 
participation.
    (3) This section has no application to any employee alleging 
discrimination prohibited by this section who, acting without direction 
from his or her employer (or the employer's agent), deliberately causes 
a violation of any requirement of the Energy Reorganization Act of 
1974, as amended, or the Act.
    (b) Any employee who believes that they have been discharged or 
otherwise discriminated against by any person for engaging in protected 
activities specified in paragraph (a)(1) of this section may seek a 
remedy for the discharge or discrimination through an administrative 
proceeding in the Department of Labor. The administrative proceeding 
must be initiated within 180 days after an alleged violation occurs. 
The employee may do this by filing a complaint alleging the violation 
with the Department of Labor, Wage and Hour Division. The Department of 
Labor may order reinstatement, back pay, and compensatory damages.
    (c) A violation of paragraph (a), (e), or (f) of this section by a 
Commission licensee, a holder of a standard design approval, an 
applicant for a Commission license, standard design certification, or a 
standard design approval, or a contractor or subcontractor of a 
Commission licensee, holder of a standard design approval, or any 
applicant may be grounds for--
    (1) Denial, revocation, or suspension of the license or standard 
design approval;
    (2) Withdrawal or revocation of a proposed or final standard design 
certification;
    (3) Imposition of a civil penalty on the licensee, holder of a 
standard design approval, or applicant (including an applicant for a 
standard design certification under this part following Commission 
adoption of final design certification rule) or a contractor or 
subcontractor of the licensee, holder of a standard design approval, or 
applicant; or
    (4) Other enforcement action.
    (d) Actions taken by an employer, or others, which adversely affect 
an employee may be predicated upon nondiscriminatory grounds. The 
prohibition applies when the adverse action occurs because the employee 
has engaged in protected activities. An employee's engagement in 
protected activities does not automatically render him or her immune 
from discharge or discipline for legitimate reasons or from adverse 
action dictated by nonprohibited considerations.
    (e)(1) Each holder or applicant for a license or design approval, 
must prominently post the revision of NRC Form 3, ``Notice to 
Employees,'' referenced in Sec.  19.11(e)(1) of this chapter. This form 
must be posted at locations sufficient to permit employees protected by 
this section to observe a copy on the way to or from their place of 
work. Premises must be posted no later than 30 days after an 
application is docketed and remain posted while the application is 
pending before the Commission, during the term of the license, and for 
30 days following license termination.
    (2) Copies of NRC Form 3 may be obtained by writing to the Regional 
Administrator of the appropriate NRC Regional Office listed in appendix 
D to 10 CFR part 20, via email to [email protected], or by 
visiting the NRC's online library at http://www.nrc.gov/reading-rm/doc-collections/forms/.
    (f) No agreement affecting the compensation, terms, conditions, or 
privileges of employment, including an agreement to settle a complaint 
filed by an employee with the Department of Labor pursuant to section 
211 of the Energy Reorganization Act of 1974, as amended, may contain 
any provision which would prohibit, restrict, or otherwise discourage 
an employee from participating in protected activity as defined in 
paragraph (a)(1) of this section including, but not limited to, 
providing information to the NRC or to his or her employer on potential 
violations or other matters within NRC's regulatory responsibilities.
    (g) Part 19 of 10 CFR sets forth requirements and regulatory 
provisions applicable to licensees, holders of a standard design 
approval, applicants for a license, standard design certification, or 
standard design approval, and contractors or subcontractors of a 
Commission licensee, or holder of a standard design approval, and are 
in addition to the requirements in this section.

[[Page 87049]]

Sec.  53.070  Completeness and accuracy of information.

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


Sec.  53.080  Specific exemptions.

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


Sec.  53.090  Standards for review.

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

[[Page 87050]]

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


Sec.  53.100  Jurisdictional limits.

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


Sec.  53.110  Attacks and destructive acts.

    Licensees, applicants for licenses, permits, certifications, and 
design approvals, and applicants for an amendment to any license, 
permit, certification, or design approval under this part are not 
required to provide for design features or other measures for the 
specific purpose of protection against the effects of--
    (a) Attacks and destructive acts, including sabotage, directed 
against the facility by an enemy of the United States, whether a 
foreign government or other person; or
    (b) Use or deployment of weapons incident to U.S. defense 
activities.


Sec.  53.115  Rights related to special nuclear material.

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


Sec.  53.117  License suspension and rights of recapture.

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


Sec.  53.120  Information collection requirements: OMB approval.

    (a) The NRC has submitted the information collection requirements 
contained in this part to the Office of Management and Budget (OMB) for 
approval as required by the Paperwork Reduction Act (44 U.S.C. 3501 et 
seq.). The NRC may not conduct or sponsor, and a person is not required 
to respond to, a collection of information unless it displays a 
currently valid OMB control number. OMB has approved the information 
collection requirements contained in this part under control number 
3150-XXXX.
    (b) The approved information collection requirements contained in 
this part appear in Sec. Sec.  53.070, 53.080, 53.240, 53.410, 53.420, 
53.425, 53.430, 53.440, 53.450, 53.480, 53.500, 53.540, 53.605, 53.610, 
53.620, 53.700, 53.710, 53.715, 53.720, 53.730, 53.780, 53.785, 53.805, 
53.810, 53.815, 53.830, 53.850, 53.855, 53.865, 53.870, 53.875, 53.880, 
53.910, 53.1010, 53.1020, 53.1030, 53.1045, 53.1060, 53.1070, 53.1075, 
53.1080, 53.1100, 53.1109, 53.1115, 53.1130, 53.1140, 53.1144, 53.1146, 
53.1173, 53. 1182, 53.1188, 53.1200, 53.1206, 53.1209, 53.1210, 
53.1221, 53.1230, 53.1236, 53.1239, 53.1241, 53.1254, 53.1257, 53,1263, 
53.1270, 53.1276, 53.1279, 53.1282, 53.1288, 53.1295, 53.1300, 53.1306, 
53.1309, 53.1312, 53.1327, 53.1330, 53.1333, 53.1336, 53.1348, 53.1360, 
53.1366, 53.1369, 53.1372, 53.1384, 53.1410, 53.1413, 53.1416, 53.1419, 
53.1437, 53.1449, 53.1452, 53.1458, 53.1470, 53.1505, 53.1510, 53.1515, 
53.1525, 53.1530, 53.1535, 53.1540, 53.1545, 53.1550, 53.1560, 53.1565, 
53.1570, 53.1575, 53.1580, 53.1620, 53.1630, 53.1645, 53.1680, 53.1690, 
53.1720.
    (c) This part contains information collection requirements in 
addition to those approved under the control number specified in 
paragraph (a) of this section. The information collection requirement 
and the control numbers under which it is approved are as follows:
    (1) In Sec. Sec.  53.765, 53.770, 53.780, and 53.795, NRC Form 396 
is approved under control number 3150-0024.
    (2) In Sec. Sec.  53.775 and 53.795, NRC Form 398 is approved under 
control number 3150-0090.
    (3) In Sec.  53.1640, NRC Form 366 is approved under control number 
3150-0104.
    (4) In Sec.  53.1630, NRC Form 361 is approved under control number 
3150-0238.
    (5) In Sec.  53.1650, International Atomic Energy Agency Design 
Information Questionnaire forms are approved under control number 3150-
0056.
    (6) In Sec.  53.1650, DOC/NRC Form AP-A and associated forms are 
approved under control numbers 0694-0135.

Subpart B--Technology-Inclusive Safety Requirements


Sec.  53.210  Safety criteria for design-basis accidents.

    Design features and programmatic controls must be provided for each 
commercial nuclear plant such that identification and analyses of 
design-basis accidents (DBAs) in accordance with Sec.  53.240 
demonstrate the following:
    (a) An individual located at any point on the boundary of the 
exclusion area for any 2-hour period following the onset of the 
postulated fission product release would not receive a radiation dose 
in excess of 25 rem (250 millisieverts) total effective dose equivalent 
(TEDE); and
    (b) An individual located at any point on the outer boundary of the 
low-population zone who is exposed to the radioactive cloud resulting 
from the postulated fission product release (during the entire period 
of its passage) would not receive a radiation dose in

[[Page 87051]]

excess of 25 rem (250 millisieverts) TEDE.\1\

    \1\ The use of 25 rem TEDE is not intended to imply that this 
number constitutes an acceptable limit for an emergency dose to the 
public under accident conditions. Rather, this dose value has been 
set forth in this section as a reference value, which can be used in 
the evaluation of plant design features with respect to postulated 
reactor accidents, to assure that these designs provide assurance of 
low risk of public exposure to radiation, in the event of an 
accident.


Sec.  53.220  Safety criteria for licensing-basis events other than 
design-basis accidents.

    Design features and programmatic controls must be provided for each 
commercial nuclear plant such that identification and analysis of 
licensing-basis events (LBEs) other than DBAs in accordance with Sec.  
53.240 demonstrate the following:
    (a) Plant SSCs, personnel, and programs provide the necessary 
capabilities and maintain the necessary reliability to address LBEs 
other than DBAs in accordance with Sec. Sec.  53.240 and 53.450(e), and 
provide measures for defense in depth in accordance with Sec.  53.250; 
and
    (b) The analysis of risks to public health and safety resulting 
from LBEs other than DBAs under Sec.  53.450(e) includes comprehensive 
risk metrics that satisfy associated risk performance objectives that 
are acceptable to the NRC and provide an appropriate level of safety.


Sec.  53.230  Safety functions.

    (a) The primary safety function is limiting the release of 
radioactive materials from the facility and must be maintained during 
normal operation and for LBEs over the life of the plant.
    (b) Additional safety functions needed to support the retention of 
radioactive materials during LBEs--such as controlling reactivity, heat 
generation, heat removal, and chemical interactions--must be identified 
for each commercial nuclear plant.
    (c) The primary and additional safety functions are required to 
satisfy the safety criteria defined in Sec. Sec.  53.210 and 53.220, or 
more restrictive alternative criteria adopted under Sec.  53.470, and 
must be fulfilled by the design features, human actions, and 
programmatic controls specified throughout this part.


Sec.  53.240  Licensing-basis events.

    (a) Licensing-basis events must be identified for each commercial 
nuclear plant and analyzed under Sec.  53.450 to demonstrate that the 
safety requirements in this subpart have been satisfied.
    (b) The identified LBEs, ranging from anticipated event sequences 
to very unlikely event sequences, must collectively address 
combinations of malfunctions of plant SSCs, human errors, facility 
hazards, and the effects of external hazards.
    (c) The analysis of LBEs must--
    (1) Include analysis of one or more DBAs under Sec.  53.450(f);
    (2) Confirm the adequacy of design features and programmatic 
controls needed to satisfy the safety criteria defined in Sec. Sec.  
53.210 and 53.220, or more restrictive alternative criteria adopted 
under Sec.  53.470, and
    (3) Establish related functional requirements for plant SSCs, 
personnel, and programs.


Sec.  53.250  Defense in depth.

    (a) Measures must be taken for each commercial nuclear plant to 
ensure appropriate defense in depth is provided to compensate for 
uncertainties in the analysis of the safety criteria such that there is 
reasonable assurance that the safety criteria in this subpart are met 
over the life of the plant.
    (b) The uncertainties that must be addressed under paragraph (a) of 
this section include those related to the state of knowledge and 
modeling capabilities, the ability of barriers to limit the release of 
radioactive materials from the facility during LBEs other than DBAs, 
the reliability and performance of plant SSCs and personnel, and the 
effectiveness of programmatic controls.
    (c) The safety analysis may not rely upon a single engineered 
design feature, human action, or programmatic control, no matter how 
robust, to address the range of LBEs other than DBAs.


Sec.  53.260  Normal operations.

    Holders of licenses to operate commercial nuclear plants under this 
part must control public doses and dose rates in unrestricted areas 
from normal plant operations to meet the requirements in 10 CFR part 
20.


Sec.  53.270  Protection of plant workers.

    Holders of licenses to operate commercial nuclear plants under this 
part must control occupational doses to meet the requirements in 10 CFR 
part 20.

Subpart C--Design and Analysis Requirements


Sec.  53.400  Design features for licensing-basis events.

    (a) Design features must be provided for each commercial nuclear 
plant such that, when combined with corresponding human actions and 
programmatic controls, the plant will satisfy the safety criteria 
defined in Sec. Sec.  53.210 and 53.220, or more restrictive 
alternative criteria adopted under Sec.  53.470.
    (b) Design features must ensure that the safety functions 
identified in Sec.  53.230 are fulfilled during licensing-basis events 
(LBEs).


Sec.  53.410  Functional design criteria for design-basis accidents.

    (a) Functional design criteria must be defined for each design 
feature required by Sec.  53.400 and relied upon to demonstrate 
compliance with the safety criteria defined in Sec.  53.210.
    (b) Corresponding human actions and programmatic controls must be 
identified and implemented in accordance with this and other subparts 
to achieve and maintain the reliability and capability of structures, 
systems, and components (SSCs) relied upon to satisfy the defined 
functional design criteria and the safety criteria required in Sec.  
53.210, and to maintain consistency with analyses required by Sec.  
53.450(f).


Sec.  53.415  Protection against external hazards.

    Safety-related (SR) SSCs must be protected against or must be 
designed to withstand the effects of natural phenomena (e.g., 
earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches) and 
constructed hazards (e.g., dams, transportation routes, military and 
industrial facilities) considering an event severity up to the design-
basis external hazard levels as determined under Sec.  53.510 without 
losing the capability to perform the safety functions identified under 
Sec.  53.230. Specific requirements for earthquake engineering are 
included in Sec.  53.480.


Sec.  53.420  Functional design criteria for licensing-basis events 
other than design-basis accidents.

    (a) Functional design criteria must be defined for each design 
feature required by Sec.  53.400 and relied upon to--
    (1) Demonstrate compliance with the safety criteria in Sec.  53.220 
or more restrictive alternative criteria adopted under Sec.  53.470; 
and
    (2) Demonstrate compliance with the evaluation criteria in Sec.  
53.450(e) or more restrictive alternative criteria adopted under Sec.  
53.470.
    (b) Corresponding human actions and programmatic controls must be 
identified and implemented in accordance with this and other subparts 
to achieve and maintain the reliability and capability of SSCs relied 
upon to--

[[Page 87052]]

    (1) Satisfy the safety criteria in Sec.  53.220 or more restrictive 
alternative criteria adopted under Sec.  53.470; and
    (2) Satisfy the evaluation criteria in Sec.  53.450(e) or more 
restrictive alternate criteria adopted under Sec.  53.470.


Sec.  53.425  Design features and functional design criteria for normal 
operations.

    (a) Design features must be provided for each commercial nuclear 
plant to support the Radiation Protection Program required in Sec.  
53.850.
    (b) Functional design criteria must be defined for each design 
feature relied upon to demonstrate compliance with Sec.  53.850.
    (c) Functional design criteria, including design objectives for 
dose to the maximally exposed member of the public, must be defined for 
design features to show that plant design features and corresponding 
programmatic controls, including monitoring programs, control liquid, 
gaseous, and solid wastes, as required under part 20 of this 
chapter.\1\

    \1\ A guide for keeping doses to the public as low as is 
reasonably achievable is that the estimated annual dose to the 
maximally exposed member of the public does not exceed 10 mrem total 
effective dose equivalent. A design objective of maintaining doses 
below 10 mrem/year should not be construed as a radiation protection 
standard.


Sec.  53.430  Design features and functional design criteria for 
protection of plant workers.

    (a) Design features must be provided for each commercial nuclear 
plant such that, when combined with corresponding programmatic 
controls, the requirements in Sec.  53.270 can be met.
    (b) Functional design criteria must be defined for each design 
feature relied upon to demonstrate compliance with Sec.  53.270.


Sec.  53.440  Design requirements.

    (a)(1) Analysis, appropriate test programs, prototype testing, 
operating experience, or a combination thereof must demonstrate that 
each design feature required by Sec.  53.400 meets the defined 
functional design criteria required by Sec. Sec.  53.410 and 53.420. 
This demonstration must consider interdependent effects throughout the 
commercial nuclear plant and the range of conditions under which the 
design features required by Sec.  53.400 must function throughout the 
plant's lifetime.
    (2) The design processes for SR and non-safety-related but safety-
significant (NSRSS) SSCs under this part must include administrative 
procedures for evaluating operating, design, and construction 
experience and for considering applicable important industry 
experiences in the design of those SSCs.
    (b) The design features required by Sec.  53.400 must, wherever 
applicable, be designed using generally accepted consensus codes and 
standards that have been endorsed or otherwise found acceptable by the 
U.S. Nuclear Regulatory Commission (NRC).
    (c) The materials used for each SR and NSRSS SSC must be qualified 
for their service conditions over the design life of the SSC.
    (d) Possible degradation mechanisms related to aging, fatigue, 
chemical interactions, operating temperatures, effects of irradiation, 
and other environmental factors that may affect the performance of SR 
and NSRSS SSCs must be evaluated and used to inform the design and the 
development of integrity assessment programs under Sec.  53.870.
    (e)(1) Safety-related and NSRSS SSCs must be designed and located 
to minimize, consistent with other safety requirements in this part, 
the probability and effect of fires and explosions.
    (2) Noncombustible and fire-resistant materials must be used 
wherever practical throughout the facility, particularly in locations 
with SR and NSRSS SSCs.
    (3) Fire detection and fire suppression systems of appropriate 
capacity and capability must be provided and designed to minimize the 
adverse effects of fires on SR and NSRSS SSCs.
    (4) Fire suppression systems must be designed to ensure that their 
rupture or inadvertent operation does not significantly impair the 
ability of SR and NSRSS SSCs to perform their safety functions to 
satisfy Sec.  53.230.
    (f) Safety and security must be considered together in the design 
process such that, where possible, security issues are effectively 
resolved through design and engineered security features.
    (g) The reactor system and waste stores for each commercial nuclear 
plant must be capable of achieving and maintaining a subcritical 
condition during normal operations and following any LBE identified in 
accordance with Sec.  53.240.
    (h) Each commercial nuclear plant must have a capability to provide 
long-term cooling of the reactor fuel and waste stores during normal 
operations and following any LBE identified in accordance with Sec.  
53.240.
    (i) The design, analysis, staffing, and programmatic controls for 
each commercial nuclear plant must consider the number of reactors, 
waste stores, and other significant inventories of radioactive 
materials and the associated operating configurations, common systems, 
system interfaces, and system interactions.
    (j)(1) Design features must be provided and related functional 
design criteria defined such that, with limited use of operator 
actions, one or more physical barriers are maintained to limit the 
release of radionuclides from reactor systems, waste stores, or other 
significant inventories of radioactive materials assuming the impact of 
a large, commercial aircraft.
    (2) The functional design criteria for those design features 
provided to address the requirements in paragraph (j)(1) of this 
section must be based on an assessment of the impact of a large, 
commercial aircraft used for long distance flights in the United 
States, with aviation fuel loading typically used in such flights, and 
an impact speed and angle of impact considering the ability of both 
experienced and inexperienced pilots to control large, commercial 
aircraft at low altitude representative of a commercial nuclear plant's 
low profile.\1\

    \1\ Changes to the detailed parameters on aircraft impact 
characteristics set forth in guidance must be approved by the 
Commission.

    (k) Design features and related functional design criteria must be 
defined such that analyses demonstrate a low risk of permanent injury 
to the public due to the health effects of the chemical hazards of 
licensed material.
    (l) Measures must be taken during the design of commercial nuclear 
plants to minimize, to the extent practicable, contamination of the 
facility and the environment, facilitate eventual decommissioning, and 
minimize, to the extent practicable, the generation of radioactive 
waste in accordance with Sec.  20.1406 of this chapter.
    (m)(1) Each commercial nuclear plant must include criticality 
monitoring capabilities meeting the requirements of either Sec.  70.24 
of this chapter or paragraph (m)(2) of this section.
    (2) In lieu of maintaining a monitoring system capable of detecting 
criticality as described in Sec.  70.24 of this chapter, criticality 
accident requirements may be satisfied by--
    (i) Demonstrating the sub-criticality of special nuclear material, 
except when it is inside the reactor and the reactor is being operated, 
by maintaining k-effective below 0.95 at a 95 percent probability, 95 
percent confidence level, under conditions that maximize reactivity for 
the applicable storage and handling configurations, and
    (ii) Providing radiation monitors for fuel storage and associated 
handling

[[Page 87053]]

areas when fuel is present to detect excessive radiation levels and to 
support initiating appropriate safety actions.
    (3) While a spent fuel transportation package approved under 10 CFR 
part 71 of this chapter or spent fuel storage cask approved under 10 
CFR part 72 is in the special nuclear material handing or storage area, 
the requirements in 10 CFR parts 71 or 72, as applicable, and the 
requirements of the certificate of compliance for that package or cask, 
are the applicable requirements for the fuel within that package or 
cask.
    (n)(1) The design of each commercial nuclear plant must reflect 
state-of-the-art human factors principles for safe and reliable 
performance in all locations that human activities are expected for 
performing or supporting the continued availability of plant safety or 
emergency response functions.
    (2) The design must provide for the capabilities described in Sec.  
53.730(b) to ensure the plant staff are able to monitor plant 
conditions and respond to events.
    (3) The means by which the design and human actions together will 
achieve the safety requirements of subpart B of this part must be 
evaluated and used to inform the design and the development of the 
concept of operations required by Sec.  53.730(c).
    (4) A functional requirements analysis and function allocation must 
be used to ensure that plant design features address how safety 
functions and functional safety criteria are satisfied, and how the 
safety functions will be assigned to appropriate combinations of human 
action, automation, active safety features, passive safety features, or 
inherent safety characteristics.


Sec.  53.450  Analysis requirements.

    (a) Requirement to have a probabilistic risk assessment (PRA). A 
PRA of each commercial nuclear plant must be performed to identify 
potential failures, susceptibility to internal and external hazards, 
and other contributing factors to event sequences that might challenge 
the safety functions identified in Sec.  53.230 and to support 
demonstrating that each commercial nuclear plant meets the safety 
criteria of Sec.  53.220, or more restrictive alternative criteria 
adopted under Sec.  53.470.
    (b) Specific uses of analyses. The PRA in combination with other 
generally accepted approaches for systematically evaluating engineered 
systems must be used--
    (1) In informing the selection of the LBEs, as described in Sec.  
53.240, which must be considered in the design to determine compliance 
with the safety criteria in subpart B of this part.
    (2) For informing the classification of SSCs according to their 
safety significance in accordance with Sec.  53.460 and for identifying 
the environmental conditions under which the SSCs and operating staff 
must perform their safety functions.
    (3) In evaluating the adequacy of defense-in-depth measures 
required in accordance with Sec.  53.250.
    (4) To identify and assess all plant operating states where there 
is the potential for the uncontrolled release of radioactive material 
to the environment.
    (5) To identify and assess events that challenge plant control and 
safety systems whose failure could lead to the uncontrolled release of 
radioactive material to the environment. These include internal events, 
such as human errors and equipment failures, and external events 
identified in accordance with subpart D of this part.
    (c) Maintenance and upgrade of analyses. The PRA must be maintained 
at least every 5 years until the permanent cessation of operations 
under Sec.  53.1070 and upgraded in conformance with generally accepted 
methods, standards, and practices that have been endorsed or otherwise 
found acceptable by the NRC.
    (d) Qualification of analytical codes. The analytical codes used in 
modeling plant behavior in analyses of licensing-basis events 
(including but not limited to thermodynamics, reactor physics, fuel 
performance, and mechanistic source term codes) must be qualified for 
the range of conditions for which they are to be used.
    (e) Analyses of licensing-basis events other than design-basis 
accidents.
    (1) Analyses must be performed for LBEs other than design-basis 
accidents (DBAs). These LBEs must be identified using insights from a 
PRA in combination with other generally accepted approaches for 
systematically evaluating engineered systems to identify and analyze 
equipment failures and human errors.
    (2) The analysis of LBEs other than DBAs must include definition of 
evaluation criteria for each event or specific categories of LBEs to 
determine the acceptability of the plant response to the challenges 
posed by internal and external hazards to provide an appropriate level 
of safety.
    (3) The analyses of LBEs other than DBAs must address event 
sequences from initiation to a defined end state and be used in 
combination with other engineering analyses to demonstrate that the 
functional design criteria required by Sec.  53.420 provide sufficient 
barriers to the unplanned release of radionuclides to satisfy the 
evaluation criteria defined for each LBE other than DBAs, to satisfy 
the safety criteria specified in accordance with Sec.  53.220 and 
provide defense in depth as required by Sec.  53.250.
    (4) The methodology used to identify, categorize, and analyze LBEs 
must include a means to identify event sequences deemed significant for 
controlling the risks posed to public health and safety.
    (f) Analysis of design-basis accidents. (1) The analysis of LBEs 
required by Sec.  53.240 must include analysis of DBAs that address 
possible challenges to the safety functions identified under Sec.  
53.230. The events selected as DBAs must be those that, if not 
terminated, have the potential for exceeding the safety criteria in 
Sec.  53.210.
    (2) The DBAs selected must be analyzed using deterministic methods 
that address event sequences from initiation to a safe stable end state 
and assume only the SR SSCs identified under Sec.  53.460 and human 
actions addressed by the requirements of subpart F of this part are 
available to perform the safety functions identified in accordance with 
Sec.  53.230.
    (3) The analysis must conservatively demonstrate compliance with 
the safety criteria in Sec.  53.210.
    (g) Other required analyses. Analyses must be performed to assess--
    (1) Fire protection. Fire protection measures to demonstrate, 
through inclusion of fires in the analysis of LBEs or by separate 
analyses, that a fire or explosion in any plant area would not--
    (i) Prevent equipment from fulfilling the safety functions 
identified in accordance with Sec.  53.230, or
    (ii) Challenge the safety criteria in Sec. Sec.  53.210 and 53.220.
    (2) Aircraft impact. Measures provided to protect against aircraft 
impacts under Sec.  53.440(j).
    (3) Dose to members of the public. Measures taken under Sec.  
53.425, including estimating--
    (i) The quantity of each of the principal radionuclides expected to 
be released annually to unrestricted areas in liquid effluents produced 
during normal reactor operations and the dose to the maximally exposed 
member of the public in unrestricted areas.
    (ii) The quantities of each of the principal radionuclides of the 
gases, halides, and particulates expected to be released annually to 
unrestricted areas in gaseous effluents produced during normal reactor 
operations and the dose to the maximally exposed member of the public 
in unrestricted areas.
    (iii) The annual external radiation dose in unrestricted areas and 
the maximally exposed member of the

[[Page 87054]]

public in unrestricted areas due to direct radiation from contained 
radiation sources from the commercial nuclear plant during normal 
reactor operations.


Sec.  53.460  Safety categorization and special treatments.

    (a) Structures, systems, and components must be classified 
according to their safety significance. The SSC categories must include 
``Safety-Related,'' ``Non-Safety-Related but Safety-Significant,'' and 
``Non-Safety-Significant,'' as defined in subpart A of this part.
    (b) For SR and NSRSS SSCs, the conditions under which they must 
perform their safety function in Sec.  53.230 must be identified. 
Special treatments must be established in accordance with this and 
other subparts to provide confidence that the SSCs will perform under 
the service conditions and with reliability consistent with the 
analysis performed under Sec.  53.450 to demonstrate meeting the safety 
criteria in Sec. Sec.  53.210 and 53.220, or more restrictive 
alternative criteria adopted under Sec.  53.470.
    (1) The special treatments for SR SSCs must include meeting the 
applicable quality assurance requirements from appendix B of part 50 of 
this chapter.
    (2) The special treatments for NSRSS SSCs and special treatments 
for SR SSCs beyond those required under (b)(1) of this section may 
include meeting selected quality assurance requirements from appendix B 
of part 50 of this chapter when such treatment is needed to address 
performance requirements, equipment reliability, or uncertainties.
    (c) Human actions needed to prevent or mitigate LBEs must be 
identified, be able to be performed reliably under the postulated 
environmental conditions, and be addressed by programs established in 
accordance with subpart F of this part to provide confidence that those 
actions will be performed as assumed in the analysis performed in 
accordance with Sec.  53.450 to demonstrate meeting the criteria in 
Sec. Sec.  53.210, 53.220, and 53.450(e), or more restrictive 
alternative criteria adopted under Sec.  53.470.


Sec.  53.470  Maintaining analytical safety margins used to justify 
operational flexibilities.

    Where an applicant or licensee so chooses, alternative criteria 
more restrictive than those defined in Sec. Sec.  53.220 and 53.450(e) 
may be adopted to support operational flexibilities. In such cases, 
applicants and licensees must ensure that the functional design 
criteria of Sec.  53.420, the analysis requirements of Sec.  53.450(e), 
and identification of special treatment of SSCs and human actions under 
Sec.  53.460 reflect and support the use of alternative criteria to 
justify operational flexibilities. Licensees must ensure that measures 
taken to provide the analytical margins supporting operational 
flexibilities are incorporated into design features and programmatic 
controls and are maintained within programs required in other subparts.


Sec.  53.480  Earthquake engineering.

    (a) Effects of earthquakes. Structures, systems, and components 
classified as SR or NSRSS must be able to withstand the effects of 
earthquakes, commensurate with the safety significance of the SSC, 
without loss of capability to perform their role in fulfilling the 
safety functions required by Sec.  53.230.
    (b) Definitions. For the purpose of this section--
    Design-Basis Ground Motions (DBGMs) are the vibratory ground 
motions for which certain SSCs must be designed to remain functional.
    Operating basis earthquake (OBE) ground motion is the vibratory 
ground motion for which those features of the commercial nuclear plant 
necessary for continued operation without undue risk to the health and 
safety of the public are designed to remain functional. The OBE ground 
motion is used in Sec.  53.720.
    Response spectrum is a plot of the maximum responses (acceleration, 
velocity, or displacement) of idealized single-degree-of-freedom 
oscillators as a function of the natural frequencies of the oscillators 
for a given damping value. The response spectrum is calculated for a 
specified vibratory motion input at the oscillators' supports.
    Surface deformation is the distortion of geologic strata on or near 
the ground surface that occurs because of tectonic forces that result 
from earthquakes.
    (c) Design considerations--(1) Design-Basis Ground Motions. (i) The 
DBGMs must be derived from the Site Ground Motion Response Spectra 
developed in accordance with Sec.  53.510(c), by taking into 
consideration the functional design criteria of SSCs in accordance with 
Sec. Sec.  53.410 and 53.420. The horizontal component of the DBGM(s) 
in the free-field at the foundation level of the structures must be an 
appropriate response spectrum that is determined based on the risk-
significance of SSCs and their safety functions. In view of the limited 
data available on vibratory ground motion of strong earthquakes, it is 
acceptable that the design response spectra be smoothed spectra.
    (ii) The commercial nuclear plant must be designed so that, if the 
DBGMs occur, the following SSCs remain functional and within applicable 
stress, strain, and deformation limits:
    (A) Structures, systems, and components for which functional design 
criteria are established in accordance with Sec.  53.410 or Sec.  
53.420; and
    (B) Structures, systems, and components classified as SR or NSRSS 
commensurate with safety significance in accordance with Sec.  53.460.
    (iii) In addition to seismic loads, applicable concurrent normal 
operating, functional, and accident-induced loads must be taken into 
account in the design of the SR SSCs and, commensurate with safety 
significance, NSRSS SSCs.
    (iv) The design of the commercial nuclear plant must take into 
account the possible effects of seismic-induced ground disruption, such 
as fissuring, lateral spreads, differential settlement, liquefaction, 
and landsliding, on the facility foundations.
    (v) The SSCs fulfilling the safety functions required by Sec.  
53.230 must be demonstrated through design, testing, or qualification 
methods to be able to fulfill those safety functions during and after 
the vibratory ground motion associated with the DBGMs.
    (vi) The evaluation of SSCs required by this section to show they 
are able to function during and after earthquake ground motion must 
take into account soil-structure interaction effects and the expected 
duration of vibratory motion. It is permissible to design for strain 
limits in excess of yield strain in some of these SSCs during the DBGMs 
and under the postulated concurrent loads, provided the necessary 
safety functions are maintained.
    (2) OBE Ground Motion. The OBE Ground Motion must be characterized 
by response spectra. The value of the OBE Ground Motion must be set to 
one-third or less of the DBGMs response spectra.
    (3) [Reserved]
    (4) Required seismic instrumentation. Suitable instrumentation must 
be provided so that the seismic response of commercial nuclear plant SR 
SSCs or NSRSS SSCs can be evaluated promptly after an earthquake.
    (d) Surface deformation. (1) The potential for surface deformation 
must be taken into account in the design of the commercial nuclear 
plant by providing reasonable assurance that in the event of 
deformation, SSCs classified as SR or NSRSS in accordance with Sec.  
53.460 will remain functional.
    (2) In addition to surface deformation induced loads, the design of 
SSCs must take into account, commensurate with

[[Page 87055]]

safety significance, seismic loads and applicable concurrent functional 
and accident-induced loads.
    (3) The design provisions for surface deformation must be based on 
its postulated occurrence in any direction and azimuth and under any 
part of the commercial nuclear plant, unless evidence indicates this 
assumption is not appropriate, and must take into account the estimated 
rate at which the surface deformation may occur.
    (e) Seismically induced floods and water waves and other design 
conditions. Seismically induced floods and water waves from either 
locally or distantly generated seismic activity and other design 
conditions determined pursuant to subpart D of this part must be taken 
into account in the design of the commercial nuclear plant so as to 
prevent undue risk to the health and safety of the public.
    (f) Analysis. The analyses required by Sec.  53.450 must address 
seismic hazards and related SSC responses in determining that the 
safety criteria defined in Sec.  53.220 will be met.
    (g) Design criteria, human actions, and programmatic controls. 
Functional design criteria, human actions, and programmatic controls 
needed to address seismic events must be identified and implemented in 
accordance with this and other subparts to achieve and maintain the 
performance of SSCs relied upon to satisfy the safety criteria in Sec.  
53.220 and to maintain consistency with analyses required by Sec.  
53.450 when accounting for the site-specific frequencies and magnitudes 
of earthquakes for a commercial nuclear plant.

Subpart D--Siting Requirements


Sec.  53.500  General siting and siting assessment.

    (a) The siting of each commercial nuclear plant must be supported 
by assessments of proposed sites such that the design, including design 
features and programmatic controls corresponding to the site 
characteristics, satisfies the safety criteria defined in Sec. Sec.  
53.210 and 53.220 or more restrictive alternative criteria adopted 
under Sec.  53.470. The siting assessment must ensure that site 
characteristics that might contribute to the initiation, progression, 
or consequences of licensing-basis events (LBEs) analyzed under 
Sec. Sec.  53.450 and 53.480 are identified and mitigated by design 
features or programmatic controls. The siting assessment must take into 
consideration the potential adverse impacts that a commercial nuclear 
plant may have on nearby populations as a result of normal operations 
or LBEs.
    (b) Activities performed to identify site characteristics or 
otherwise needed to determine site-specific contributors to functional 
design criteria or analysis assumptions under subpart C of this part 
must satisfy the applicable special treatment requirements of Sec.  
53.460, including, where applicable, the quality assurance requirements 
from appendix B of part 50 of this chapter.


Sec.  53.510  External hazards.

    (a) General external hazard requirements. The design-basis external 
hazard level for the relevant external hazards for a site must be 
identified and characterized based on site-specific assessments of 
natural and constructed hazards with the potential to adversely affect 
plant functions. The external hazard frequencies and magnitudes 
determined from the site-specific assessments must take into account 
uncertainties and variabilities in data, models, and methods relied on 
to characterize the external hazards.
    (b) Definitions. For the purpose of this section, the following 
terms mean:
    Geological siting factors are geological and seismic factors that 
may affect the design and operation of the proposed commercial nuclear 
plant.
    Ground Motion Response Spectra (GMRS) are the site-specific GMRS 
resulting from the geologic investigations and evaluations of the site 
vicinity and region and used to determine design-basis ground motions 
for structures, systems, and components under Sec.  53.480.
    Probabilistic seismic hazard analysis is an analytical methodology 
that incorporates uncertainty into estimates of an annual frequency of 
exceedance for a certain ground motion parameter (e.g., peak ground 
acceleration, peak ground velocity, response spectral values) at a 
site.
    (c) Geological investigations. The GMRS for the site must be 
determined based on the results of investigations of the geological, 
seismological, and engineering characteristics of the site and its 
environs and must be characterized by both horizontal and vertical 
free-field GMRS at the free ground surface. The size of the region to 
be investigated and the type of data pertinent to the investigations 
must be determined based on the nature of the region surrounding the 
site. Data on vibratory ground motion, earthquake recurrence rates, 
fault geometry and slip rates, and site subsurface material properties 
must be obtained by reviewing pertinent literature and carrying out 
field investigations. Uncertainties are inherent in the parameters and 
models used to estimate the GMRS for the site. The site assessment must 
reflect these uncertainties through an appropriate analysis, such as a 
probabilistic seismic hazard analysis.
    (d) Geologic and seismic siting factors. The geologic and seismic 
siting factors considered for design under Sec. Sec.  53.415 and 53.480 
must include, but are not limited to, determination of the potential 
for surface tectonic and nontectonic deformations, the size and 
character of seismically induced floods and water waves that could 
affect a site from either locally or distantly generated seismic 
activity, soil and rock stability, liquefaction potential, and natural 
and artificial slope stability.


Sec.  53.520  Site characteristics.

    Site characteristics that might contribute to the initiation, 
progression, or consequences of LBEs analyzed under Sec.  53.450 must 
be identified, assessed, and considered in the design and analyses 
required by subpart C of this part.


Sec.  53.530  Population-related considerations.

    Every site must have an exclusion area, a low-population zone, and 
a population center distance as defined in Sec.  53.020.
    (a) The offsite radiological consequences estimated by the analyses 
required by Sec.  53.450(f) must be used to confirm that--
    (1) An individual located at any point on the boundary of the 
exclusion area for any 2-hour period following onset of the postulated 
fission product release would not receive a radiation dose in excess of 
25 rem (250 millisieverts) total effective dose equivalent.
    (2) An individual located at any point on the outer boundary of the 
low-population zone who is exposed to the radioactive cloud resulting 
from the postulated fission product release (during the entire period 
of its passage) would not receive a radiation dose in excess of 25 rem 
(250 millisieverts) total effective dose equivalent.
    (b) The population center distance must be at least one and one-
third times the distance from the reactor to the outer boundary of the 
low-population zone. The boundary of the population center must be 
determined upon consideration of population distribution. Political 
boundaries are not controlling in the calculation of population center 
distance.
    (c) Reactor sites should be located away from very densely 
populated centers. Areas of low-population density

[[Page 87056]]

are, generally, preferred. However, in determining the acceptability of 
a particular site located away from a very densely populated center but 
not in an area of low-population density, consideration will be given 
to safety, environmental, economic, or other factors, which may result 
in the site being found acceptable.


Sec.  53.540  Siting interfaces.

    Site characteristics must be addressed by the design features, 
programmatic controls, and supporting analyses used to demonstrate that 
the safety criteria in Sec. Sec.  53.210 and 53.220 are met for each 
commercial nuclear plant. Site characteristics must be such that 
adequate emergency plans and security plans can be developed and 
maintained.

Subpart E--Construction and Manufacturing Requirements


Sec.  53.600  Construction and manufacturing--scope and purpose.

    This subpart applies to those construction and manufacturing 
activities authorized by a construction permit (CP), combined license 
(COL), manufacturing license (ML), or limited work authorization (LWA) 
issued under this part.


Sec.  53.605  Reporting of defects and noncompliance.

    Each CP and ML issued under this part is subject to the terms and 
conditions in this section, and each COL issued under this part is 
subject to the terms and conditions in this section until the date that 
the Commission makes the finding under Sec.  53.1452(g).
    (a) Definitions. The definitions in Sec.  21.3 of this chapter 
apply to this section.
    (b) Posting requirements. (1) Each individual, partnership, 
corporation, dedicating entity, or other entity subject to the 
regulations in this section must post current copies of this section 
and the regulations in 10 CFR part 21; section 206 of the Energy 
Reorganization Act of 1974, as amended; and procedures adopted under 
these regulations. These documents must be posted in a conspicuous 
position on any premises within the United States where the activities 
subject to the license are conducted.
    (2) If posting of these regulations or the procedures adopted under 
them is not practical, the licensee may, in addition to posting section 
206 of the Energy Reorganization Act of 1974, as amended, post a notice 
that describes the regulations/procedures, including the name of the 
individual to whom reports may be made, and states where they may be 
examined.
    (c) Procedures. The holder of a CP, COL, or ML subject to this 
section must adopt appropriate procedures to--
    (1) Evaluate deviations and failures to comply to identify defects 
and failures to comply associated with substantial safety hazards as 
soon as practicable, and, except as provided in paragraph (c)(2) of 
this section, in all cases within 60 days of discovery, to identify a 
reportable defect or failure to comply that could create a substantial 
safety hazard, were it to remain uncorrected.
    (2) Ensure that if an evaluation of an identified deviation or 
failure to comply potentially associated with a substantial safety 
hazard cannot be completed within 60 days from the discovery of the 
deviation or failure to comply, an interim report is prepared and 
submitted to the Commission through a director or responsible officer, 
or designated person as discussed in paragraph (d)(5) of this section. 
The interim report should describe the deviation or failure to comply 
that is being evaluated and should also state when the evaluation will 
be completed. This interim report must be submitted in writing within 
60 days of discovery of the deviation or failure to comply.
    (3) Ensure that a director or responsible officer of the holder of 
a CP, COL, or ML subject to this section is informed as soon as 
practicable, and, in all cases, within the 5 working days after 
completion of the evaluation described in paragraph (c)(1) or (c)(2) of 
this section, if the construction or manufacture of a facility or 
activity, or a basic component supplied for such a facility or 
activity--
    (i) Fails to comply with the Atomic Energy Act of 1954, as amended, 
or any applicable regulation, order, or license of the Commission 
relating to a substantial safety hazard;
    (ii) Contains a defect; or
    (iii) Underwent any significant breakdown in any portion of the 
quality assurance program (QAP) conducted under the requirements of 
appendix B to part 50 of this chapter that could have produced a defect 
in a basic component. These breakdowns in the QAP are reportable 
whether or not the breakdown actually resulted in a defect in a design 
approved and released for construction, installation, or manufacture.
    (d) Reporting defects and noncompliance. (1) The holder of a CP, 
COL, or ML subject to this section that obtains information reasonably 
indicating that the facility or manufactured reactors fails to comply 
with the Atomic Energy Act of 1954, as amended, or any applicable 
regulation, order, or license of the Commission relating to a 
substantial safety hazard must notify the Commission of the failure to 
comply through a director, responsible officer, or designated person as 
discussed in paragraph (d)(5) of this section.
    (2) The holder of a CP, COL, or ML subject to this section that 
obtains information reasonably indicating the existence of any defect 
found in the construction or manufacture, or any defect found in the 
final design of a facility as approved and released for construction or 
manufacture, must notify the Commission of the defect through a 
director, responsible officer, or designated person as discussed in 
paragraph (d)(5) of this section.
    (3) The holder of a CP, COL, or ML subject to this part, who 
obtains information reasonably indicating that the QAP has undergone 
any significant breakdown discussed in paragraph (c)(3)(iii) of this 
section must notify the Commission of the breakdown in the QAP through 
a director, responsible officer, or designated person as discussed in 
paragraph (d)(5) of this section.
    (4) When acting as a dedicating entity, the holder of a CP, COL, or 
ML subject to this section is responsible for identifying and 
evaluating deviations; reporting defects and failures to comply 
associated with substantial safety hazards for dedicated items; and 
maintaining auditable records for the dedication process.
    (5) The notification requirements of this paragraph apply to all 
defects and failures to comply associated with a substantial safety 
hazard regardless of whether extensive evaluation, redesign, or repair 
is required to conform to the criteria and bases stated in the Safety 
Analysis Report, CP, COL, or ML. Evaluation of potential defects and 
failures to comply and reporting of defects and failures to comply 
under this section satisfies the CP holder's, COL holder's, and ML 
holder's evaluation and notification obligations under 10 CFR part 21, 
and satisfies the responsibility of individual directors or responsible 
officers or holders of a CP, COL, or ML subject to this section to 
report defects, and failures to comply associated with substantial 
safety hazards under section 206 of the Energy Reorganization Act of 
1974, as amended. The director or responsible officer may authorize an 
individual to provide the notification required by this section. 
However, this does not relieve the director or responsible officer of 
his or her responsibility under this section.

[[Page 87057]]

    (e) Notification--timing and where sent. The notification required 
by paragraph (d) of this section must consist of--
    (1) Initial notification by telephone, facsimile, or email 
identified in appendix A to 10 CFR part 73 to the U.S. Nuclear 
Regulatory Commission (NRC) Operations Center within 2 days following 
receipt of information by the director or responsible corporate officer 
under paragraph (c)(3) of this section, on the identification of a 
defect or a failure to comply. If the CP, COL, or ML holder elects to 
use facsimile, verification that the facsimile has been received should 
be made by calling the NRC Operations Center. This paragraph does not 
apply to interim reports described in paragraph (c)(2) of this section.
    (2) Written notification submitted to the NRC Document Control Desk 
by an appropriate method listed in Sec.  53.040, with a copy to the 
appropriate NRC Regional Administrator at the address specified in 
appendix D to 10 CFR part 20 and a copy to the appropriate NRC resident 
inspector, if applicable, within 30 days following receipt of 
information by the director or responsible corporate officer under 
paragraph (c)(3) of this section, on the identification of a defect or 
failure to comply.
    (f) Content of notification. The written notification required by 
paragraph (e)(2) of this section must clearly indicate that the written 
notification is being submitted under this section and include the 
following information, to the extent known.
    (1) Name and address of the individual or individuals informing the 
Commission.
    (2) Identification of the facility, the activity, or the basic 
component supplied for the facility or the activity within the United 
States which contains a defect or fails to comply.
    (3) Identification of the firm constructing or manufacturing the 
facility or supplying the basic component which fails to comply or 
contains a defect.
    (4) Nature of the defect or failure to comply and the safety hazard 
which is created or could be created by the defect or failure to 
comply.
    (5) The date on which the information of a defect or failure to 
comply was obtained.
    (6) In the case of a basic component that contains a defect or 
failure to comply, the number and location of these components in use 
at the facility subject to the regulations in this part.
    (7) In the case of a completed reactor manufactured under this 
part, the entities to which the reactor was supplied.
    (8) The corrective action which has been, is being, or will be 
taken; the name of the individual or organization responsible for the 
action; and the length of time that has been or will be taken to 
complete the action.
    (9) Any advice related to the defect or failure to comply about the 
facility, activity, or basic component that has been, is being, or will 
be given to other entities.
    (g) Procurement documents. Each holder of a CP, COL, or ML subject 
to this section must ensure that each procurement document for a 
facility or a basic component specifies the provisions of 10 CFR part 
21 or this section that apply, as applicable.
    (h) Coordination with 10 CFR part 21. The requirements of this 
section are satisfied when the defect or failure to comply associated 
with a substantial safety hazard has been previously reported under 10 
CFR part 21, under Sec.  73.1205 of this chapter, under this section, 
or under Sec.  53.1640.
    (i) Records retention. The holder of a CP, COL, or ML subject to 
this section must prepare and maintain records necessary to accomplish 
the purposes of this section, specifically--
    (1) Retain procurement documents, which define the requirements 
that facilities or basic components must satisfy in order to be 
considered acceptable, for the lifetime of the facility or basic 
component.
    (2) Retain records of evaluations of all deviations and failures to 
comply under paragraph (c)(1) of this section for the longest of--
    (i) Ten years from the date of the evaluation;
    (ii) Five years from the date that an early site permit is 
referenced in an application for a COL; or
    (iii) Five years from the date of delivery of a manufactured 
reactor.
    (3) Retain records of all interim reports to the Commission made 
under paragraph (c)(2) of this section, or notifications to the 
Commission made under paragraph (d) of this section for the minimum 
time periods stated in paragraph (i)(2) of this section;
    (4) Suppliers of basic components must retain records of--
    (i) All notifications sent to affected licensees or purchasers 
under paragraph (d)(4) of this section for a minimum of 10 years 
following the date of the notification;
    (ii) The facilities or other purchasers to whom the basic 
components or associated services were supplied for a minimum of 15 
years from the delivery of the basic component or associated services.
    (5) Maintaining reports in accordance with this section satisfies 
the recordkeeping obligations under 10 CFR part 21 of the entities, 
including directors or responsible officers thereof, subject to this 
section.


Sec.  53.610  Construction.

    (a) Management and control. Licensees must ensure that the 
following plans, programs, and organizational units are developed and 
implemented to manage and control the construction activities:
    (1) Programs to ensure that the construction of a commercial 
nuclear plant supports the eventual compliance with the design and 
analysis requirements in subpart C of this part.
    (2) An organization, headed by qualified personnel, responsible for 
managing, controlling, and evaluating the adequacy of the construction 
activities.
    (3) Procedures describing the qualifications for personnel in key 
positions in the licensee's management and control organization and the 
organizational responsibilities, authority, and interfaces with other 
parts of the licensee's organization.
    (4) Procedures to evaluate the applicability of other national and 
international construction experience to the planned and ongoing 
construction activities and to ensure the applicable experience will be 
provided to those constructing the plant.
    (5) A fitness-for-duty program, under 10 CFR part 26.
    (6)(i) A QAP meeting the requirements of appendix B of part 50 of 
this chapter as required by Sec.  53.460(b).
    (ii) Appropriate programmatic controls to provide special treatment 
for non-safety-related but safety-significant structures, systems, and 
components (SSCs).
    (7) A radiation protection program, in accordance with 10 CFR part 
20, that includes measures for monitoring the dose to individuals 
working with radioactive materials brought onto the site, as 
applicable.
    (8) An information security program in accordance with Sec. Sec.  
73.21, 73.22, and 73.23 of this chapter, as applicable.
    (b) Construction activities. No person may begin the construction 
of a commercial nuclear plant on a site on which the facility is to be 
operated under this part until that person has been issued either a CP 
or COL, an early site permit authorizing activities under Sec.  
53.1130, or an LWA under this part.
    (1) Licensees must satisfy the following requirements:

[[Page 87058]]

    (i) As appropriate, considering the types and quantities of 
radioactive materials being brought onto the site--
    (A) The licensee must maintain and follow a special nuclear 
material (SNM) material control and accounting program, a measurement 
control program, and other material control procedures that include 
corresponding record management requirements as required by the 
provisions of Sec.  70.32 of this chapter. Prior to initial receipt of 
SNM onsite, the licensee must implement an SNM material control and 
accounting program in accordance with 10 CFR part 74.
    (B) Procedures must be in place to receive, possess, use, and store 
source, byproduct, and SNM in accordance with applicable portions of 10 
CFR parts 30, 40, and 70.
    (C) A plant staff training program associated with the receipt of 
radioactive material must be approved and implemented prior to initial 
receipt of byproduct, source or SNM (excluding exempt quantities as 
described in Sec.  30.18 of this chapter).
    (ii) For construction of a commercial nuclear plant involving 
multiple reactor units, plans and procedures must be in place to 
prevent or mitigate potential hazards to the SSCs of operating units 
resulting from construction activities, including the managerial and 
administrative controls to be used to provide assurance that the 
limiting conditions for operation of the operating units are not 
exceeded as a result of construction activities.
    (iii) Procedures must be in place prior to the start of 
construction activities that describe how construction will be 
controlled so as not to impact other features important to the design, 
such as dewatering, slope stability, backfill, compaction, and seepage.
    (iv) For LWA holders, a plan must be developed for redress of 
activities performed under the LWA should one of the following 
situations arise:
    (A) LWA work activities are terminated by the holder of the LWA;
    (B) The LWA is revoked by the NRC; or
    (C) The Commission denies the associated CP or COL application.
    (2)(i) Onsite fresh fuel must be protected and stored in compliance 
with Sec.  73.67 of this chapter.
    (ii) Before initial fuel load into the reactor (or, for a fueled 
manufactured reactor, before initiating the physical removal of any one 
of the independent physical mechanisms to prevent criticality required 
under Sec.  53.620(d)(1)), a cybersecurity program that meets the 
requirements of Sec. Sec.  73.54 or 73.110 of this chapter, a physical 
security program that meets the requirements of Sec. Sec.  73.55 or 
73.100 of this chapter, and an access authorization program that meets 
the requirements of Sec. Sec.  73.56 or 73.120 of this chapter must be 
established, as applicable.
    (iii) Fire protection measures must be implemented for work and 
storage areas (including adjacent fire areas that could affect the work 
or storage area) before initial receipt of byproduct, source, or non-
fuel SNM (excluding exempt quantities as described in Sec.  30.18 of 
this chapter). The fire protection measures for areas associated with 
new fuel (including all fuel handling, fuel storage, and adjacent fire 
areas that could affect the new fuel) must be implemented before 
receipt of fuel. Prior to the receipt of fuel, a formal letter of 
agreement must be in place with the local fire department specifying 
the nature of arrangements in support of the fire protection program.
    (c) Inspection and acceptance. (1) The licensee must have a process 
for accepting individual or groups of SSCs upon completion of 
construction and protecting them from damage or tampering as other 
construction activities continue.
    (2) The post construction acceptance process must address the 
inspections, tests, analyses, and acceptance criteria specified in the 
COL under Sec.  53.1440 or the equivalent verifications needed to 
support the issuance of an operating license under Sec.  53.1387.


Sec.  53.620  Manufacturing.

    (a) Management and control. Holders of MLs must ensure that the 
following plans, programs, and organizational units are developed and 
implemented to manage and control the manufacturing activities within 
the scope of the ML:
    (1) Programs to ensure that the manufacturing of a manufactured 
reactor or portions of a manufactured reactor complies with the design 
and analysis requirements in subpart C of this part. The entity with 
design authority for the manufactured reactor covered by the ML must be 
identified in the license.
    (2) An organizational and management structure responsible for 
managing, controlling, and evaluating the adequacy of the reactor 
design and manufacturing activities.
    (3) Procedures describing the qualifications for personnel in key 
positions in the licensee's management and control organization and the 
organizational responsibilities, authority, and interfaces with other 
parts of the licensee's organization.
    (4) A program to evaluate the applicability of other national and 
international design and manufacturing experience to the planned and 
ongoing manufacturing activities.
    (5) A fitness-for-duty program, in accordance with 10 CFR part 26.
    (6)(i) A QAP meeting the requirements of appendix B to part 50 of 
this chapter, to be applied to the design, fabrication, construction, 
and testing of the SSCs of the manufactured reactor.
    (ii) Appropriate programmatic controls to provide special treatment 
measures for non-safety-related but safety-significant SSCs.
    (7) A radiation protection program, in accordance with 10 CFR part 
20, that includes measures for monitoring the dose to individuals if 
the manufacturing activities include working with radioactive 
materials.
    (8) An information security program in accordance with Sec. Sec.  
73.21, 73.22 and 73.23 of this chapter, as applicable.
    (b) Manufacturing activities. Holders of MLs must satisfy the 
following requirements:
    (1) The manufacturing process must be conducted within facilities 
for which the ML holder has the authority to establish controls on any 
activity that might affect manufacturing. The licensee must establish 
access controls to the portions of each facility involved in the 
manufacturing processes governed by the ML.
    (2) Manufacturing processes must be performed in accordance with 
the ML and the referenced codes and standards that have been endorsed 
or otherwise found acceptable by the NRC.
    (3) A post-manufacturing inspection and acceptance process must be 
established and implemented before transporting a manufactured reactor 
or portions of a manufactured reactor for installation at a commercial 
nuclear plant. The process must consider the results of inspections, 
tests, and analyses that have been performed and the acceptance 
criteria that are necessary and sufficient to conclude that 
manufacturing activities have been completed in accordance with the ML.
    (c) Control of radioactive materials. As appropriate considering 
the types and quantities of radioactive materials being brought into 
the manufacturing facility--
    (1) Procedures must be in place to receive, transfer, possess, and 
use source, byproduct, and SNM in accordance with the applicable 
portions of 10 CFR parts 30, 40 and 70.
    (2) A fire protection program must be established and implemented 
before the initial receipt of byproduct, source, or

[[Page 87059]]

non-fuel SNM (excluding exempt quantities as described in Sec.  30.18 
of this chapter).
    (3) An emergency plan appropriate for responding to the facility-
specific hazards of an accidental release of radioactive material and 
to limit the health effects of the associated chemical hazards of 
licensed material must be approved and implemented prior to the receipt 
of byproduct, source, or SNM (excluding exempt quantities as described 
in Sec.  30.18 of this chapter).
    (4) A plant staff training program associated with the receipt of 
radioactive material must be approved and implemented before initial 
receipt of byproduct, source, or SNM (excluding exempt quantities as 
described in Sec.  30.18 of this chapter).
    (5) Security requirements must be implemented for the protection of 
SNM based on the type, enrichment, and quantity in accordance with 10 
CFR part 73, as applicable, and for the protection of Category 1 and 
Category 2 quantities of radioactive material in accordance with 10 CFR 
part 37, as applicable.
    (d) Fuel loading. (1)(i) An ML may authorize possession of a 
manufactured reactor into which the licensee has loaded fresh 
(unirradiated) fuel pursuant to a license issued under part 70 of this 
chapter only if the manufactured reactor is configured during its 
loading, storage, and transport with at least two independent physical 
mechanisms in place, each of which is sufficient to prevent criticality 
assuming optimum neutron moderation and neutron reflection conditions.
    (ii) The ML applicant may file a separate, subsequent application 
for the 10 CFR part 70 license or combine the application for the 10 
CFR part 70 license with the application for an ML.
    (iii) The Commission has determined that any such fueled 
manufactured reactor in which the independent physical mechanisms to 
prevent criticality have been installed is not in operation.
    (iv) Upon installation of the fueled manufactured reactor in its 
place of operation and a Commission finding that the acceptance 
criteria in the COL that authorized reactor construction are met under 
Sec.  53.1452(g), the independent physical mechanisms to prevent 
criticality may be removed. Upon initiating the physical removal of any 
one of the independent physical mechanisms to prevent criticality, the 
fueled manufactured reactor has commenced operation.
    (2) Holders of 10 CFR part 70 licenses authorizing the possession 
and loading of fresh fuel into manufactured reactors must comply with 
the requirements of 10 CFR part 70 for the facilities and activities 
related to the storage, movement, and loading of fresh fuel in the 
manufactured reactor. Holders of these 10 CFR part 70 licenses must 
comply with the requirements of Subpart H to 10 CFR part 70, regardless 
of whether their proposed activities meet the applicability criteria 
found in 10 CFR 70.60. Procedures, equipment, and personnel required by 
the 10 CFR part 70 license, must be in place before the receipt of SNM 
at the manufacturing facility.
    (i) Before the receipt of SNM, the licensee must have security 
programs in place that meet the performance objectives of 10 CFR 73.67, 
with the following additions and exceptions:
    (A) A physical security plan describing the physical security 
program must be maintained and a cybersecurity program must be 
established for the possession and loading of fresh fuel into a 
manufactured reactor authorized by a 10 CFR part 70 license, regardless 
of fuel type, enrichment, and quantity.
    (B) The physical security program must be designed to prevent 
unintended and uncontrolled criticality events.
    (C) The cybersecurity program must provide reasonable assurance 
that a cyberattack would not adversely impact the functions performed 
by digital assets used by the licensee for implementing the physical 
security requirements of this section, or the radiation monitoring and 
criticality requirements in this section or in 10 CFR part 70.
    (D) All holders of a part 70 license that authorizes loading of 
fresh fuel into a manufactured reactor must perform the screening 
required in Sec.  73.67(d)(4) of this chapter to confirm the identity, 
trustworthiness, and reliability of individuals prior to granting 
unescorted access to special nuclear material; these determinations 
must be documented.
    (ii) [Reserved]
    (3) The loading or unloading of fresh fuel into or from a 
manufactured reactor and any changes to the configuration of reactivity 
control and prevention systems for the fueled manufactured reactor must 
be performed by a certified fuel handler meeting the requirements in 
subpart F of this part.
    (e) Transportation. (1) A holder of an ML may not transport or 
allow to be removed from the places of manufacture the manufactured 
reactor or portions thereof as defined in the ML except for transport 
to a site for which the Commission has issued a COL that references the 
subject ML.
    (2) A holder of an ML must include in any contract governing the 
transport of a manufactured reactor or portions thereof as defined in 
the ML from the places of manufacture to any other location, a 
provision requiring that the person transporting the manufactured 
reactor comply with all shipping requirements in applicable NRC 
regulations, certificates of compliance, and NRC-issued licenses.
    (3) Procedures governing the preparation of the manufactured 
reactor or portions thereof as defined in the ML for transport and the 
conduct of the transport must be issued prior to transport. The 
procedures must implement the protective measures and restrictions 
described in NRC regulations and NRC-issued licenses to protect the 
reactor from potential conditions that would adversely affect the safe 
operation of a commercial nuclear plant.
    (4) For a manufactured reactor that is to be loaded with fresh fuel 
before transport to the place of operation, the ML must specify that 
transportation will be in accordance with parts 71 and 73 of this 
chapter.
    (f) Acceptance and installation at the site for which the 
Commission has issued a COL that references the subject ML. (1) 
Installation at the site for which the Commission has issued a COL that 
references the subject ML must follow the regulations in Sec.  53.610.
    (2) Upon arrival at the site, the manufactured reactor or portions 
of a manufactured reactor may not be installed in its place of 
operation unless the COL holder performs inspections sufficient to 
verify the reactor is in compliance with the ML and has not been 
damaged in transit. The COL holder must perform these inspections in 
accordance with documented procedures subject to quality assurance 
measures commensurate with their importance to safety. In addition, 
inspections must confirm that the interface requirements between the 
manufactured reactor or portions of a manufactured reactor and the 
remaining portions of the commercial nuclear plant are met.

Subpart F--Requirements for Operation


Sec.  53.700  Operational objectives.

    (a) Each holder of an operating license (OL) or combined license 
(COL) under this part must develop, implement, and maintain controls 
for plant structures, systems, and components (SSCs), responsibilities 
of plant personnel, and plant programs during the operating life of 
each commercial nuclear plant such that the requirements defined in 
subpart B are satisfied. More specifically:

[[Page 87060]]

    (1) Each holder of an OL or COL under this part must maintain the 
capabilities, availability, and reliability of plant SSCs to ensure 
that the safety functions identified in Sec.  53.230 will be performed 
if called upon during licensing-basis events (LBEs).
    (2) Each holder of an OL or COL under this part must ensure that 
plant personnel have adequate knowledge and skills to perform their 
assigned duties that support the performance of the safety functions 
identified in Sec.  53.230.
    (3) Each holder of an OL or COL under this part must implement 
plant programs sufficient to ensure that the safety functions 
identified in Sec.  53.230 will be performed if called upon during 
normal operations and LBEs.
    (b) [Reserved]


Sec.  53.710  Maintaining capabilities and availability of structures, 
systems, and components.

    Controls must be provided for each commercial nuclear plant 
licensed under this part such that the capabilities, availability, and 
reliability of plant SSCs, when combined with corresponding 
programmatic controls and human actions, provide that the safety 
criteria defined in Sec. Sec.  53.210 and 53.220 will be met.
    (a) Technical specifications must be developed, implemented, and 
maintained that define conditions or limitations on plant operations 
that are necessary to ensure that safety-related (SR) SSCs can fulfill 
the safety functions identified under Sec.  53.230 and support meeting 
the safety criteria of Sec.  53.210. The technical specifications must 
describe the following requirements:
    (1) Limits on the inventory of radioactive materials within the 
reactor system and supporting systems with the potential, individually 
or collectively, to cause a release exceeding the safety criteria in 
Sec.  53.210 as a result of a design-basis accident analyzed in 
accordance with Sec.  53.450(f).
    (2) Operating limits for the facility that if exceeded could lead 
to a failure to perform a required safety function necessary to 
demonstrate compliance with the safety criteria in Sec.  53.210.
    (3) For each SSC classified as SR in accordance with Sec.  53.460, 
technical specifications must define--
    (i) Limiting conditions for operation. Limiting conditions for 
operation are the lowest functional capability or performance levels of 
SR SSCs required to ensure that the design-basis accidents analyzed in 
accordance with Sec.  53.450(f) satisfy the safety criteria of Sec.  
53.210. When a limiting condition for operation is not met, the 
licensee must shut down the plant or follow any remedial action 
permitted by the technical specifications until the condition can be 
met.
    (ii) Surveillance requirements. Surveillance requirements are 
requirements relating to test, calibration, or inspection to assure 
that the necessary quality of systems and components is maintained and 
that the limiting conditions for operation will be met.
    (4) Design elements to be included are those elements of the plant 
such as materials of construction and geometric arrangements, which, if 
altered or modified, would have a significant effect on safety and are 
not covered in categories described in paragraphs (a)(1) through (3) of 
this section.
    (5) Administrative controls are the provisions relating to 
organization and management, procedures, recordkeeping, review and 
audit, and reporting necessary to assure operation of the plant in a 
safe manner. Each licensee must submit any reports to the Commission 
pursuant to approved technical specifications under Sec.  53.040.
    (b) Controls on plant operations, including availability controls, 
must be developed and implemented to ensure that the configurations and 
special treatments for SR SSCs and non-safety-related but safety-
significant (NSRSS) SSCs provide the capabilities, availability, and 
reliability required to demonstrate compliance with the criteria of 
Sec. Sec.  53.220 and 53.450(e).\1\ The controls must--

    \1\ The comprehensive risk metrics and related risk performance 
objectives established under Sec.  53.220 involve assessing and 
averaging the risks over a defined period (e.g., plant year) and do 
not constitute a real-time requirement that must be continuously 
demonstrated by the licensee.

    (1)(i) Identify who within the commercial nuclear plant has 
authority to make configuration changes;
    (ii) Establish processes to make configuration changes to NSRSS 
SSCs; and
    (iii) Establish processes to ensure that all organizations of the 
commercial nuclear plant affected by the configuration changes are 
formally notified and approve of the change.
    (2) Describe how the special treatments for each NSRSS SSC and 
special treatments for SR SSCs beyond those under paragraph (a) of this 
section will be established and maintained over the operating life of 
the commercial nuclear plant.


Sec.  53.715  Maintenance, repair, and inspection programs.

    (a) A program to control maintenance activities and monitor the 
performance or condition of SR and NSRSS SSCs must be developed, 
implemented, and maintained.
    (b) Whenever a licensee determines through activities related to 
maintenance, repair, and inspection of SSCs, the activities under Sec.  
53.710, or otherwise that the performance or condition of an SR or 
NSRSS SSC does not demonstrate compliance with established special 
treatments or performance goals related to capabilities, availability, 
or reliability, the licensee must take appropriate corrective action.
    (c) Performance and condition monitoring activities and associated 
goals and preventive maintenance activities must be evaluated at least 
every 24 months. The evaluations must take into account, where 
practical, industry-wide operating experience. Adjustments must be made 
where necessary to ensure that the objective of preventing failures of 
SSCs through maintenance is appropriately balanced against the 
objective of minimizing unavailability of SSCs due to monitoring or 
preventive maintenance.
    (d) Before performing maintenance activities (including but not 
limited to surveillance, post-maintenance testing, and corrective and 
preventive maintenance), the licensee must assess and manage the 
increase in risk that may result from the proposed maintenance 
activities.


Sec.  53.720  Response to seismic events.

    If vibratory ground motion exceeding that of the operating basis 
earthquake ground motion or significant plant damage due to vibratory 
ground motion occurs, the licensee must shut down the commercial 
nuclear plant. If structures, systems, or components necessary for the 
safe shutdown of the commercial nuclear plant are not available after 
the occurrence of this vibratory ground motion, the licensee must 
consult with the Commission and must propose a plan for the timely, 
safe shutdown of the commercial nuclear plant. Prior to resuming 
operations, the licensee must demonstrate to the Commission that those 
features necessary for continued operation without undue risk to the 
health and safety of the public or necessary to maintain the licensing 
basis of the commercial nuclear plant were either not functionally 
damaged or have been repaired.


Sec.  53.725  General staffing, training, personnel qualifications, and 
human factors requirements.

    (a) Two classes of commercial nuclear plants. Commercial nuclear 
plants licensed under this part are either of the

[[Page 87061]]

class, based upon the similarity of operating and technical 
characteristics of the plants in the class, of self-reliant-mitigation 
facilities or of interaction-dependent-mitigation facilities. A 
commercial nuclear plant is a self-reliant-mitigation facility if the 
U.S. Nuclear Regulatory Commission (NRC) determined as part of its 
approval of the OL or COL for that plant that its design demonstrates 
compliance with the criteria of Sec.  53.800(a)(1) through (a)(5). 
Otherwise, the commercial nuclear plant is an interaction-dependent-
mitigation facility.
    (b) Purpose and applicability. The regulations in Sec. Sec.  53.725 
through 53.830 address areas related to staffing, training, personnel 
qualifications, and human factors engineering for applicants for or 
holders of OLs or COLs under this part. These regulations are organized 
as follows:
    (1) Sections 53.725 through 53.745 address general requirements for 
staffing, training, personnel qualifications, and human factors 
engineering. The regulations within these sections are applicable to 
all applicants for or holders of OLs or COLs under this part, except 
where specifically stated otherwise.
    (2) Sections 53.760 through 53.795 address operator and senior 
operator licensing requirements. The regulations within these sections 
are applicable to those applicants for or holders of OLs or COLs under 
this part for interaction-dependent-mitigation facilities that have not 
yet certified the permanent cessation of operations and permanent 
removal of fuel from the reactor vessel as described under Sec.  
53.1070.
    (3) Sections 53.800 through 53.820 address generally licensed 
reactor operator requirements. The regulations within these sections 
are in lieu of Sec. Sec.  53.760 through 53.795 for those applicants 
for or holders of OLs or COLs under this part for self-reliant-
mitigation facilities that have not yet certified the permanent 
cessation of operations and permanent removal of fuel from the reactor 
vessel as described under Sec.  53.1070.
    (4) Section 53.830 provides general personnel training 
requirements. The regulations within this section are applicable to all 
applicants for or holders of OLs or COLs under this part.
    (c) Definitions. When used in Sec. Sec.  53.725 through 53.830:
    Applicant refers to an applicant for an operator or senior operator 
license; licensee refers to the holder of an operator, senior operator, 
or generally licensed reactor operator license; and facility licensee 
refers to the licensee for the commercial nuclear plant where the 
applicant would be licensed or the licensee is licensed.
    Automation means a device or system that accomplishes (partially or 
fully) a function or task.
    Auxiliary operator means any individual who operates components of 
a commercial nuclear plant but does not manipulate controls or direct 
the manipulation of controls of the plant and is not required to be 
licensed under the provisions of this part.
    Controls when used with respect to a nuclear reactor means 
apparatus and mechanisms, the manipulation of which directly affects 
the reactivity or power level of the reactor.
    Generally licensed reactor operator means any individual licensed 
under the provisions of Sec.  53.810 to manipulate controls of a self-
reliant-mitigation facility and to direct the licensed activities of 
generally licensed reactor operators.
    Interaction-dependent-mitigation facility means a commercial 
nuclear plant design other than one that demonstrates compliance with 
the operating and technical characteristics defined under Sec.  53.800.
    Load following means a commercial nuclear plant automatically 
changing its output to match expected demand in response to externally 
originated instructions or signals.
    Operator means any individual licensed under the provisions of 
Sec. Sec.  53.760 through 53.795 to manipulate controls of an 
interaction-dependent-mitigation facility.
    Performance testing means testing conducted to verify a simulation 
facility's performance as compared to actual or predicted reference 
plant performance.
    Reference plant means the specific commercial nuclear plant on 
which a simulation facility's configuration, system control 
arrangement, and design data are based. The reference plant may or may 
not be constructed.
    Self-reliant-mitigation facility means a commercial nuclear plant 
design that demonstrates compliance with the operating and technical 
characteristics defined under Sec.  53.800.
    Senior operator means any individual licensed under the provisions 
of Sec. Sec.  53.760 through 53.795 to manipulate controls of an 
interaction-dependent-mitigation facility and to direct the licensed 
activities of operators.
    Simulation facility means an interface designed to provide a 
realistic imitation of the operation of a commercial nuclear plant used 
for the administration of examinations, for training, and/or to 
demonstrate compliance with experience requirements for applicants or 
licensees. A simulation facility may rely, in whole or part, upon the 
physical utilization of the reference plant itself.
    Systems approach to training means a training program that includes 
the following five elements:
    (1) Systematic analysis of the jobs to be performed.
    (2) Learning objectives derived from the analysis which describe 
desired performance after training.
    (3) Training design and implementation based on the learning 
objectives.
    (4) Evaluation of trainee mastery of the objectives during 
training.
    (5) Evaluation and revision of the training based on the 
performance of trained personnel in the job setting.


Sec.  53.726  Communications.

    (a) An applicant or licensee or facility licensee must submit any 
communication or report required by the regulations contained within 
Sec. Sec.  53.725 through 53.830 and must submit any application filed 
under these regulations to the Commission.
    (b) Each licensee that is required to comply with the requirements 
of Sec. Sec.  53.760 through 53.795 (i.e., interaction-dependent-
mitigation facilities) must notify the appropriate NRC contact within 
30 days of the following in regard to a licensed operator or senior 
operator:
    (1) Permanent reassignment from the position for which the licensee 
has certified the need for a licensed operator or senior operator under 
Sec.  53.775(a)(1);
    (2) Termination of any operator or senior operator; or
    (3) Permanent disability or illness as required under Sec.  55.770 
of this chapter.


Sec.  53.728  Completeness and accuracy of information.

    Information provided to the Commission by an applicant for an 
operator or senior operator license or by a licensee or information 
required by statute or by the Commission's regulations, orders, or 
license conditions to be maintained by the applicant or the licensee 
must be complete and accurate in all material respects.


Sec.  53.730  Defining, fulfilling, and maintaining the role of 
personnel in ensuring safe operations.

    Each applicant for or holder of an OL or COL for a commercial 
nuclear plant under this part must comply with the following:
    (a) Human factors engineering design requirements. The plant design 
must reflect state-of-the-art human factors engineering principles for 
safe and reliable performance in all locations that

[[Page 87062]]

human activities are expected for performing or supporting the 
continued availability of plant safety or emergency response functions.
    (b) Human system interface design requirements. The plant design 
must provide for the following to support operating personnel in 
monitoring plant conditions and responding to plant events:
    (1) Features for displaying to operating personnel a minimum set of 
parameters that define the safety status of the plant and are capable 
of displaying both the full range of important plant parameters and 
data trends on demand, as well as indicating when process limits are 
being approached or exceeded;
    (2) Automatic indication of the bypassed and operable status of 
safety systems;
    (3) Direct indication of SSC status that relates to the ability of 
the SSC to perform its safety function, such as relief and safety valve 
position (i.e., open or closed) for barriers important to fulfilling 
safety functions of with such devices, and ultimate heat sink and 
cooling system status and availability;
    (4) Instrumentation to measure, record, and display key plant 
parameters related to the performance of SSCs and the integrity of 
barriers important to fulfilling safety functions to support operators 
in monitoring plant conditions and responding to plant events. Examples 
include temperatures and pressures within important systems or 
structures, core or fuel system conditions (including possible damage 
states), temperatures and levels associated with cooling functions, 
combustible gas concentrations, radiation levels in systems and within 
structures, and radioactive effluent releases;
    (5) Leakage control and detection in the design of systems that 
pass through barriers important to fulfilling safety functions for the 
release of radionuclides. An example is an SSC that penetrates a 
containment structure that might contain radioactive materials that 
could contribute to the source term during an accident;
    (6) Monitoring of in-plant radiation and airborne radioactivity as 
appropriate for a broad range of normal operating and accident 
conditions; and
    (7) For self-reliant-mitigation facilities, the plant design must 
also provide the generally licensed reactor operators with the 
capability to do the following:
    (i) Receive plant operating data, including reactor parameters and 
information needed for the evaluation of emergency conditions.
    (ii) Immediately initiate a reactor shutdown from their location.
    (iii) Promptly dispatch operations and maintenance personnel.
    (iv) Immediately implement responsibilities under the facility 
emergency plan, as applicable.
    (c) Concept of operations. A concept of operations that is of 
sufficient scope and detail to address the following must be provided:
    (1) Plant goals;
    (2) The roles and responsibilities of operating personnel and 
automation (or any combination thereof) that are responsible for 
completing plant functions;
    (3) Staffing, qualifications, and training;
    (4) The management of normal operations;
    (5) The management of off-normal conditions and emergencies;
    (6) The management of maintenance and modifications; and
    (7) The management of tests, inspections, and surveillances.
    (d) Functional requirements analysis and function allocation. A 
functional requirements analysis and a function allocation must be 
provided that are sufficient to demonstrate compliance with the 
following:
    (1) The functional requirements analysis must address how safety 
functions and functional safety criteria are satisfied, and
    (2) The function allocation must describe how the safety functions 
will be assigned to human action, automation, active safety features, 
passive safety features, and/or inherent safety characteristics.
    (e) Operating experience. A program, during construction and during 
operation, as applicable, for evaluating and applying operating 
experience must be developed, implemented, and maintained.
    (f) Staffing plan. A staffing plan must be developed and comply 
with the following:
    (1) The staffing plan must include a description of how engineering 
expertise will be available to the on-shift operating personnel during 
all plant conditions, to assist if they encounter a situation not 
covered by procedures or training. Engineering expertise includes 
familiarity with the operation of the plant for which the expertise is 
provided and one of the following:
    (i) A bachelor's degree in engineering, engineering technology, or 
physical science from an institution accredited by a U.S. government 
recognized accrediting body or equivalent; or
    (ii) A Professional Engineer's license from a U.S. State or 
territory.
    (2) Applicants for or holders of OLs or COLs for interaction-
dependent-mitigation facilities must include within their staffing 
plans a description of how the proposed numbers, positions, and 
qualifications of operators and senior operators across all modes of 
plant operations will be sufficient to ensure that plant safety 
functions will be maintained. This description must be supported by 
human factors engineering analyses and assessments.
    (3) Applicants for or holders of OLs or COLs for self-reliant-
mitigation facilities must include within their staffing plans a 
description of how generally licensed reactor operator staffing that is 
both sufficient to continually monitor the operations of fueled 
reactors and to provide for a continuity of responsibility for facility 
operations at all times during the operating phase will be maintained.
    (4) Applicants for or holders of OLs or COLs under this part must 
include within their staffing plans a description of how the numbers, 
positions, and responsibilities of personnel contained within those 
plans will adequately support all necessary functions within areas such 
as plant operations, equipment surveillance and maintenance, 
radiological protection, chemistry control, fire brigades, engineering, 
security, and emergency response.
    (5) The staffing plan must be approved by the NRC as part of its 
approval of the OL or COL for the plant. The approved staffing plan is 
subject to the requirements of Sec.  53.1565.
    (g) Training, examination, and proficiency programs. Develop, 
implement, and maintain programs that comply with the following 
requirements. These programs must be approved by the NRC as part of its 
approval of the OL or COL for the plant:
    (1) For those applicants for or holders of OLs or COLs for 
interaction-dependent-mitigation facilities:
    (i) The operator licensing initial training program required under 
Sec.  53.780(a);
    (ii) The operator licensing initial examination program required 
under Sec.  53.780(b);
    (iii) The operator licensing requalification program required under 
Sec.  53.780(c); and
    (iv) The operator proficiency program required under Sec.  
53.780(g).
    (2) For those applicants for or holders of OLs or COLs for self-
reliant-mitigation facilities, the generally licensed reactor operator 
training, examination, and proficiency programs required under Sec.  
53.815.

[[Page 87063]]

    (3) The operator licensing requalification programs required under 
Sec.  53.780(c) or Sec.  53.815(b) must be implemented upon commencing 
the administration of initial examinations under the operator licensing 
examination program required under Sec.  53.780(b) or Sec.  53.815(b), 
respectively.


Sec.  53.735  General exemptions.

    The regulations in Sec. Sec.  53.725 through 53.830 do not require 
a license for an individual who--
    (a) Under the direction and in the presence of an operator or 
senior operator or a generally licensed reactor operator, as 
appropriate, manipulates the controls of a commercial nuclear plant as 
a part of the individual's training in a facility licensee's training 
program as approved by the Commission to qualify for an operator or 
senior operator license or a generally licensed reactor operator 
license there, as appropriate, under these regulations; or
    (b) Under the direction and in the presence of a senior operator or 
generally licensed reactor operator, as appropriate, manipulates the 
controls of a commercial nuclear plant to load or unload the fuel into, 
out of, or within the reactor vessel while the reactor is not 
operating.


Sec.  53.740  Facility licensee requirements--General.

    (a) Facility licensees must demonstrate compliance with the 
requirements of either Sec. Sec.  53.760 through 53.795 for 
interaction-dependent-mitigation facilities or Sec. Sec.  53.800 
through 53.820 for self-reliant-mitigation facilities.
    (b) The facility licensee must maintain the staffing complement 
described under its approved facility staffing plan until such time as 
the permanent cessation of operations and permanent removal of fuel 
from the reactor vessel has been certified as described under Sec.  
53.1070. The approved staffing plan is subject to the requirements of 
Sec.  53.1565.
    (c) Except as provided under Sec.  53.735, the facility licensee 
may not permit the manipulation of the controls of a commercial nuclear 
plant by anyone who is not an operator or senior operator or generally 
licensed reactor operator, as appropriate.
    (d) Facility licensees for interaction-dependent-mitigation 
facilities that have not yet certified the permanent cessation of 
operations and permanent removal of fuel from the reactor vessel as 
described under Sec.  53.1070 must designate senior operators to be 
responsible for supervising the licensed activities of operators.
    (e) Apparatus and mechanisms other than controls, the operation of 
which may affect the reactivity or power level of a reactor, must be 
manipulated only while plant conditions are being monitored by an 
individual who is an operator or senior operator or a generally 
licensed reactor operator, as appropriate.
    (f)(1) Load following is permitted if at least one of the following 
is immediately capable of refusing demands when they could challenge 
the safe operation of the plant or when precluded by the plant 
equipment conditions:
    (i) The actuation of an automatic protection system that utilizes 
setpoints more conservative than those otherwise credited for the 
purposes of reactor protection; or
    (ii) An automated control system; or
    (iii) An operator or senior operator or a generally licensed 
reactor operator, as appropriate.
    (2) The provisions of paragraph (e) of this section do not apply 
during load following operations.
    (g)(1) Facility licensees for interaction-dependent-mitigation 
facilities must have present during alteration of the core (including 
fuel loading or transfer) an individual holding a senior operator 
license, or a senior operator license limited to fuel handling to 
directly supervise the activity and, during this time, the facility 
licensee must not assign other duties to this person.
    (2) Facility licensees for self-reliant-mitigation facilities must 
have present during alteration of the core (including fuel loading or 
transfer) an individual holding a generally licensed reactor operator 
license to directly supervise the activity and, during this time, the 
facility licensee must not assign other duties to this person.
    (3) The provisions of paragraphs (g)(1) and (2) of this section do 
not apply to core alterations performed as part of refueling operations 
while a facility that is capable of online refueling is operating at 
power.
    (h) Facility licensees may take reasonable action that departs from 
a license condition or a technical specification (contained in a 
license issued under this part) in an emergency when this action is 
immediately needed to protect the public health and safety and no 
action consistent with license conditions and technical specifications 
that can provide adequate or equivalent protection is immediately 
apparent. Such facility licensee action must be approved, as a minimum, 
by a senior operator or a generally licensed reactor operator, as 
applicable, or, after certifying the permanent cessation of operations 
and permanent removal of fuel from the reactor vessel as described 
under Sec.  53.1070 by a certified fuel handler, senior operator, or 
generally licensed reactor operator, as applicable, prior to taking the 
action.


Sec.  53.745  Operator license requirements.

    A person must be authorized by a license issued by the Commission 
to perform the function of an operator, senior operator, or generally 
licensed reactor operator as defined in this part.


Sec.  53.760  Operator licensing.

    (a) Applicability. Sections 53.760 through 53.795 address operator 
and senior operator licensing requirements. The regulations within 
these sections are applicable to those applicants for or holders of OLs 
or COLs under this part for interaction-dependent-mitigation facilities 
that have not yet certified the permanent cessation of operations and 
permanent removal of fuel from the reactor vessel as described under 
Sec.  53.1070.
    (b) Reserved.


Sec.  53.765  Medical requirements.

    (a) An applicant for an operator or senior operator license must 
have a medical examination by a physician. An operator or senior 
operator must have a medical examination by a physician every 2 years.
    (b) To certify the medical fitness of an applicant for an operator 
or senior operator license, an authorized representative of the 
facility licensee must complete and sign NRC Form 396, ``Certification 
of Medical Examination by Facility Licensee,'' which can be obtained by 
writing the Office of the Chief Information Officer, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, by calling 301-415-
7232, or by visiting the NRC's website at https://www.nrc.gov and 
selecting forms from the index found on the home page, or by other 
means provided by the NRC.
    (1) Form NRC 396 must certify that a physician has conducted the 
medical examination of the applicant as required in paragraph (a) of 
this section.
    (2) When the medical certification requests a conditional license 
based on medical evidence, the medical evidence must be submitted on 
NRC Form 396 to the Commission to enable the Commission to make a 
determination in accordance with Sec.  53.775(b).
    (c) The facility licensee must document and maintain the results of 
medical qualifications data, test results,

[[Page 87064]]

and each operator's or senior operator's medical history for the 
current license period and provide the documentation to the Commission 
upon request. The facility licensee must retain this documentation 
while an individual performs the functions of an operator or senior 
operator.


Sec.  53.770  Incapacitation because of disability or illness.

    If, during the term of the operator or senior operator license, the 
licensee develops a permanent physical or mental condition that causes 
the licensee to fail to demonstrate compliance with the requirements of 
Sec.  53.775(b)(1)(i), the facility licensee must notify the Commission 
within 30 days of learning of the diagnosis. For conditions for which a 
conditional license (as described in Sec.  53.775(b)) is requested, the 
facility licensee must provide medical certification on Form NRC 396 to 
the Commission (as described in Sec.  53.765(b)).


Sec.  53.775  Applications for operators and senior operators.

    (a) How to apply. (1) The applicant for an operator or senior 
operator license must--
    (i) Complete NRC Form 398, ``Personal Qualification Statement-- 
Licensee,'' which can be obtained by writing the Office of the Chief 
Information Officer, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, by calling 301-415-5877, or by visiting the NRC's website 
at https://www.nrc.gov and selecting forms from the index found on the 
home page, or by other means provided by the NRC;
    (ii) File an original of NRC Form 398, or an equivalent electronic 
submittal, together with the information required in paragraphs 
(a)(1)(iii) and (a)(1)(iv) of this section, with the appropriate 
Regional Administrator.
    (iii) Provide evidence that the applicant, as a trainee, has 
successfully demonstrated competence in manipulating the controls of 
either the facility for which a license is sought or a simulation 
facility that demonstrates compliance with the requirements of Sec.  
53.780(e). For operators applying for a senior operator license, 
certification that the operator has successfully operated the controls 
of the facility as an operator will be accepted; and
    (iv) Provide certification by the facility licensee of medical 
condition and general health on Form NRC 396, to comply with Sec.  
53.765.
    (2) The Commission may at any time after the application has been 
filed, and before the license has expired, require further information 
under oath or affirmation to enable it to determine whether to grant or 
deny the application or whether to revoke, modify, or suspend the 
license.
    (3) An applicant whose application has been denied because of a 
medical condition or their general health may submit a further medical 
report at any time as a supplement to the application.
    (4) Each application and statement must contain complete and 
accurate disclosure as to all matters required to be disclosed. The 
applicant must sign statements required by paragraphs (a)(1)(i) and 
(a)(1)(ii) of this section.
    (b) Disposition of an initial application. (1) License approval. 
The Commission will approve an initial application if it finds that the 
following criteria are met:
    (i) Health. The applicant's medical condition and general health 
will not adversely affect the performance of assigned operator or 
senior operator job duties or cause operational errors endangering 
public health and safety. The Commission will base its finding upon the 
certification by the facility licensee as detailed in Sec.  53.765(b).
    (ii) Examination. The applicant has passed the requisite 
examination in accordance with Sec.  53.780(b). The examination 
determines whether the applicant for an operator's or senior operator's 
license has learned to operate a facility competently and safely, and 
additionally, in the case of a senior operator, whether the applicant 
has learned to supervise the licensed activities of operators 
competently and safely.
    (2) Conditional license. If an applicant's general medical 
condition does not demonstrate compliance with the minimum standards 
under Sec.  53.775(b)(1)(i) of this section, the Commission may approve 
the application and include conditions in the license to accommodate 
the medical condition. The Commission will consider the recommendations 
and supporting evidence of the facility licensee and of the examining 
physician (provided on Form NRC 396) in arriving at its decision.
    (c) Re-applications. (1) An applicant whose application for a 
license has been denied because of failure to pass the examination may 
file a new application. The application must be submitted on Form NRC 
398 and include a statement signed by an authorized representative of 
the facility licensee by whom the applicant will be employed that 
states in detail the extent of the applicant's additional training and 
remediation since the denial and certifies that the applicant is ready 
for re-examination.
    (2) An applicant who has passed a portion of the examination and 
failed another may request in a new application on Form NRC 398 to be 
excused from re-examination on the portions of the examination that the 
applicant has passed. The Commission may in its discretion grant the 
request if it determines that sufficient justification is presented.


Sec.  53.780  Training, examination, and proficiency program.

    (a) Operator licensing initial training program. (1) A program that 
is based upon a systems approach to training, as defined by Sec.  
53.725(b), must be utilized for the training of applicants for operator 
and senior operator licenses. The program must ensure that applicants 
at the facility will possess the knowledge, skills, and abilities 
necessary to protect the public health and maintain those plant safety 
functions specific to the facility design. The program must be approved 
by the Commission prior to its use for training applicants, as 
described under Sec.  53.730(g). The approved operator licensing 
initial training program is subject to the requirements of Sec.  
53.1565.
    (2) The facility licensee must maintain operator licensing initial 
training program records documenting the initial operator licensing 
training administered and completed by each applicant. The facility 
licensee must retain these records during the period in which any 
trainees subsequently remain licensed as operators or senior operators 
at the facility.
    (b) Operator licensing initial examination program. (1) The 
facility licensee must establish and implement an examination program 
for testing a representative sample of the knowledge, skills, and 
abilities needed to safely perform operator and senior operator duties, 
to include both the examination methods and criteria to be used to 
assess passing performance. The program must provide for valid and 
reliable examinations and be approved by the Commission prior to its 
use for examining applicants, as described under Sec.  53.730(g). The 
approved operator licensing initial examination program is subject to 
the requirements of Sec.  53.1565.
    (2) The facility licensee must submit prepared examinations to the 
Commission for review and approval in advance of their administration.
    (3) The Commission will either administer an approved examination 
or allow the facility licensee to administer the examination. The 
facility licensee must ensure that sufficient advance notification is 
provided to the

[[Page 87065]]

Commission to either administer the examination or allow for a 
representative of the Commission to be afforded the opportunity to be 
present when the facility licensee administers the examination.
    (4) Graded examination documentation for each applicant must be 
promptly provided to the Commission for review in making operator 
licensing decisions.
    (5) The facility licensee must maintain operator licensing initial 
examination program records documenting the participation of each 
operator and senior operator applicant in the initial examination. The 
records must contain copies of examinations administered, the answers 
given by the applicant, documentation of the grading of examinations, 
and documentation of any additional training administered in areas in 
which an applicant exhibited deficiencies. The facility licensee must 
retain these records during the period in which the associated 
operators or senior operators remain licensed at the facility.
    (c) Operator licensing requalification program. (1) A program based 
upon a systems approach to training, as defined by Sec.  53.725(b), 
must be utilized for the continuing training of operators and senior 
operators.
    (i) The program must ensure that operators and senior operators at 
the facility maintain the knowledge, skills, and abilities necessary to 
protect the public health and maintain those plant safety functions 
specific to the facility design. The program must be conducted for a 
continuous period not to exceed 24 months in duration.
    (ii) The program must be approved by the Commission prior to its 
use for continuing training, as described under Sec.  53.730(g). The 
approved operator licensing requalification program is subject to the 
requirements of Sec.  53.1565.
    (2) The following requirements apply to operator licensing 
requalification programs:
    (i) The facility licensee must propose a requalification 
examination program for testing, for each requalification period, a 
sample of the topics included under the systems approach to training, 
to include both the examination methods and criteria to be used to 
assess passing performance. The program must provide for valid and 
reliable examinations and be approved by the Commission prior to its 
use for examining operators and senior operators, as described under 
Sec.  53.730(g). The approved requalification examination program is 
subject to the requirements of Sec.  53.1565.
    (ii) The following requirements apply to the requalification 
examination program:
    (A) The facility licensee must make prepared requalification 
examinations available to the Commission for review.
    (B) The facility licensee must ensure that a representative of the 
Commission is afforded the opportunity to be present during 
requalification examination administration.
    (C) The facility licensee must ensure that each operator and senior 
operator is administered a complete requalification examination on a 
periodicity not to exceed 24 months. Additionally, the facility 
licensee must ensure that any licensed operator or senior licensed 
operator who either demonstrates unsatisfactory performance on, or 
fails to complete, the biennial requalification examination is removed 
from the performance of licensed operator and senior licensed operator 
duties until such time that any necessary remedial training has been 
completed and a retake examination has been passed.
    (D) The facility licensee must promptly provide a summary of 
examination results for each operator and senior operator following the 
completion of the requalification examination.
    (3) The facility licensee must maintain operator licensing 
requalification program records documenting the participation of each 
operator and senior operator in the requalification program. The 
records must contain copies of examinations administered, the answers 
given by the operator or senior operator, documentation of the grading 
of examinations, and documentation of any additional training 
administered in areas in which an operator or senior operator exhibited 
deficiencies. The facility licensee must retain these records until the 
operator's or senior operator's license is renewed.
    (d) Examination integrity. Applicants, operators and senior 
operators, and facility licensees must not engage in any activity that 
compromises the integrity of any application or examination required by 
Sec. Sec.  53.760 through 53.795. The integrity of an examination is 
considered compromised if any activity, regardless of intent, affected, 
or, but for detection, could have affected the equitable and consistent 
administration of the examination. This includes activities related to 
the preparation and certification of applications and all activities 
related to the preparation, administration, and grading of examinations 
required by Sec. Sec.  53.760 through 53.795.
    (e) Simulation facilities. (1) This section addresses the use of a 
simulation facility for the administration of examinations, for 
training, or to demonstrate compliance with experience requirements for 
applicants for operator and senior operator licenses.
    (2) Simulation facilities used for training purposes, for 
demonstrating compliance with experience requirements, or for the 
conduct of examinations under Sec.  53.780(b) and (c) must demonstrate 
compliance with the following criteria as they relate to the facility 
licensee's reference plant:
    (i) The simulation facility must be of sufficient scope and 
fidelity for individuals to acquire and demonstrate the necessary 
knowledge, skills, and abilities to safely perform operator and senior 
operator duties.
    (ii) The simulation facility must utilize models relating to 
nuclear, thermal-hydraulic, and other applicable design-specific 
characteristics that either replicate the most recent fuel load in the 
reference commercial nuclear plant or, prior to initial fuel load (or, 
for a fueled manufactured reactor, prior to initiating the physical 
removal of any one of the independent physical mechanisms to prevent 
criticality required under Sec.  53.620(d)(1)), replicate the intended 
initial fuel load for the reference commercial nuclear plant, with the 
exception of those portions of the simulation facility that utilize the 
reference plant itself.
    (iii) Simulation facility fidelity must be demonstrated so that 
significant control manipulations are completed without procedural 
exceptions, simulator performance exceptions, or deviation from the 
approved training scenario sequence.
    (3) Facility licensees that maintain a simulation facility that has 
been approved by the Commission for training purposes, demonstrating 
compliance with experience requirements, or the conduct of examinations 
under Sec.  53.780(b) and (c) for the facility licensee's reference 
plant must:
    (i) Conduct performance testing throughout the life of the 
simulation facility in a manner sufficient to ensure that paragraph 
(e)(2) of this section is met;
    (ii) Retain the results of performance testing for 4 years after 
the completion of each performance test or until superseded by updated 
test results;
    (iii) Promptly correct modeling and hardware discrepancies and 
discrepancies identified from scenario validation and from performance 
testing or provide justification as to why the

[[Page 87066]]

presence of such discrepancies will not adversely affect simulator 
performance with respect to the criteria of paragraph (e)(2) of this 
section;
    (iv) Make the results of any uncorrected performance test failures 
that may exist at the time of the initial license examination or 
requalification examination available for NRC review, prior to or 
concurrent with preparations for each initial license examination or 
requalification examination; and
    (v) Maintain the provisions for license application and examination 
integrity consistent with Sec.  53.780(d).
    (4) A simulation facility must demonstrate compliance with the 
requirements of paragraphs (e)(2) and (e)(3) of this section for the 
Commission to accept the simulation facility for conducting initial 
examinations as described in Sec.  53.780(b), requalification training 
as described in Sec.  53.780(c), or performing control manipulations 
that affect reactivity to establish eligibility for an operator or 
senior operator license as described in Sec.  53.775(a).
    (f) Waiver of examination requirement. On application, the 
Commission may waive any or all of the requirements for an examination 
if it finds that the applicant has demonstrated the required knowledge, 
skills, and abilities to safely operate the plant, and is capable of 
continuing to do so. The Commission may make such a finding based on 
demonstration of the following:
    (1) Operating experience at a comparable facility;
    (2) Proof of the applicant's past competent and safe performance; 
and
    (3) Proof of the applicant's current qualifications.
    (g) Proficiency. The facility licensee must develop, implement, and 
maintain a proficiency program to ensure that operators and senior 
operators will actively perform the functions of an operator or senior 
operator, respectively, as needed to maintain proficiency with on-shift 
duties and familiarity with plant status. This program must include 
those steps that will be taken to re-establish proficiency when it 
cannot be maintained. This program must be approved by the Commission 
as part of its approval of the OL or COL for the plant. The approved 
proficiency program is subject to the requirements of Sec.  53.1565.
    (h) Records. Each record required by this section must be legible 
throughout the retention period specified by each Commission 
regulation. The record may be the original, a reproduced copy, or an 
electronic copy provided that the copy is authenticated by authorized 
personnel.


Sec.  53.785  Conditions of operator and senior operator licenses.

    Each operator and senior operator license contains and is subject 
to the following conditions whether stated in the license or not:
    (a) Neither the license nor any right under the license may be 
assigned or otherwise transferred.
    (b) The license is limited to the facility for which it is issued.
    (c) The license is limited to those controls of the facility or 
facilities specified in the license.
    (d) The license is subject to, and the licensee must observe, all 
applicable rules, regulations, and orders of the Commission.
    (e) The licensee must maintain or re-establish proficiency in 
accordance with the facility licensee's Commission-approved proficiency 
program required under Sec.  53.780(g).
    (f) The licensee must be subject to the facility's Commission-
approved operator licensing requalification and requalification 
examination programs required under Sec.  53.780(c).
    (g) The licensee must have a biennial medical examination as 
described by Sec.  53.765.
    (h) The licensee must notify the Commission within 30 days about a 
conviction for a felony.
    (i) The licensee must not consume or ingest alcoholic beverages 
within the protected area of commercial nuclear plants. The licensee 
must not use, possess, or sell any illegal drugs. The licensee must not 
perform activities authorized by a license issued under this part while 
under the influence of alcohol or any prescription, over-the-counter, 
or illegal substance that could adversely affect his or her ability to 
safely and competently perform his or her licensed duties. For the 
purpose of this paragraph, with respect to alcoholic beverages and 
drugs, the term ``under the influence'' means the licensee exceeded, as 
evidenced by a confirmed test result, the lower of the cutoff levels 
for drugs or alcohol contained in 10 CFR part 26, or as established by 
the facility licensee. The term ``under the influence'' also means the 
licensee could be mentally or physically impaired as a result of 
substance use including prescription and over-the-counter drugs, as 
determined under the provisions, policies, and procedures established 
by the facility licensee for its fitness-for-duty program, in such a 
manner as to adversely affect his or her ability to safely and 
competently perform licensed duties.
    (j) Each licensee must participate in the drug and alcohol testing 
programs as required under 10 CFR part 26.
    (k) The licensee must comply with any other conditions that the 
Commission may impose to protect health or to minimize danger to life 
or property.


Sec.  53.790  Issuance, modification, and revocation of operator and 
senior operator licenses.

    (a) Issuance of operator and senior operator licenses. If the 
Commission determines that an applicant for an operator license or a 
senior operator license demonstrates compliance with the requirements 
of the Atomic Energy Act of 1954, as amended, (the Act) and its 
regulations, it will issue a license in the form and containing any 
conditions and limitations it considers appropriate and necessary.
    (b) Modification and revocation of operator and senior operator 
licenses. (1) The terms and conditions of all operator and senior 
operator licenses are subject to amendment, revision, or modification 
by reason of rules, regulations, or orders issued in accordance with 
the Act or any amendments thereto.
    (2) Any license may be revoked, suspended, or modified, in whole or 
in part--
    (i) For any material false statement in the application or in any 
statement of fact required under section 182 of the Act;
    (ii) Because of conditions revealed by the application or statement 
of fact or any report, record, inspection, or other means that would 
warrant the Commission to refuse to grant a license on an original 
application;
    (iii) For willful violation of, or failure to observe, any of the 
terms and conditions of the Act or the license, or of any rule, 
regulation, or order of the Commission;
    (iv) For any conduct determined by the Commission to be a hazard to 
safe operation of the facility; or
    (v) For the sale, use, or possession of illegal drugs, or refusal 
to participate in the facility drug and alcohol testing program, or a 
confirmed positive test for drugs, drug metabolites, or alcohol in 
violation of the conditions and cutoff levels established by Sec.  
53.785(i) or the consumption of alcoholic beverages within the 
protected area of commercial nuclear plants, or a determination of 
unfitness for scheduled work as a result of the consumption of 
alcoholic beverages.


Sec.  53.795  Expiration and renewal of operator and senior operator 
licenses.

    (a) Expiration. (1) Each operator license and senior operator 
license expires 6 years after the date of

[[Page 87067]]

issuance, upon termination of employment with the facility licensee, or 
upon determination by the facility licensee that the licensed 
individual no longer needs to maintain a license.
    (2) If a licensee files an application for renewal or an upgrade of 
an existing license on Form NRC 398 at least 30 days before the 
expiration of the existing license, it does not expire until 
disposition of the application for renewal or for an upgraded license 
has been finally determined by the Commission. Filing by mail will be 
deemed to be complete at the time the application is postmarked
    (b) Renewal. (1) The applicant for renewal of an operator license 
or senior operator license must--
    (i) Complete and sign Form NRC 398 and include the number of the 
license for which renewal is sought.
    (ii) File an original of NRC Form 398 as specified in Sec.  53.775.
    (iii) Provide written evidence of the applicant's experience under 
the existing license and the approximate number of hours that the 
licensee has operated the facility.
    (iv) Provide a statement by an authorized representative of the 
facility licensee that during the effective term of the current license 
the applicant has satisfactorily completed the requalification program 
for the facility for which operator or senior operator license renewal 
is sought.
    (v) Provide evidence that the applicant has discharged the license 
responsibilities competently and safely. The Commission may accept as 
evidence of the applicant's having met this requirement a certificate 
of an authorized representative of the facility licensee or holder of 
an authorization by which the licensee has been employed.
    (vi) Provide certification by the facility licensee of medical 
condition and general health on Form NRC 396, to comply with Sec.  
53.765.
    (2) The license will be renewed if the Commission finds that--
    (i) The medical condition and the general health of the licensee 
continue to be such as not to cause operational errors that endanger 
public health and safety. The Commission will base this finding upon 
the certification by the facility licensee as described in Sec.  
53.765(b).
    (ii) The licensee--
    (A) Is capable of continuing to competently and safely assume 
licensed duties;
    (B) Has successfully completed a requalification program that has 
been approved by the Commission as required by Sec.  53.780(c); and
    (C) Has passed the requalification examinations as required by 
Sec.  53.780(c).
    (iii) There is a continued need for an operator to operate or for a 
senior operator to supervise operators at the facility designated in 
the application.
    (iv) The past performance of the licensee has been satisfactory to 
the Commission. In making its finding, the Commission will include in 
its evaluation information such as notices of violations or letters of 
reprimand in the licensee's docket.


Sec.  53.800  Facility licensees for self-reliant-mitigation 
facilities.

    (a) A commercial nuclear plant is a self-reliant-mitigation 
facility if the NRC determined as part of its approval of the OL or COL 
for that plant that its design demonstrates compliance with criteria 
(a)(1) though (a)(5) of this section. A self-reliant-mitigation 
facility is of a class, based upon the similarity of operating and 
technical characteristics of the plants in the class, such that its 
licensee must comply with the requirements of Sec. Sec.  53.800 through 
53.820 in lieu of those in Sec. Sec.  53.760 through 53.795.
    (1) The safety performance criteria of Sec. Sec.  53.210 and 53.220 
and, if applicable, any alternative criteria used in accordance with 
Sec.  53.470, must be met without reliance upon human action for 
credited event mitigation.
    (2) The results of a probabilistic risk analysis must demonstrate 
that the evaluation criteria for the events analyzed in accordance with 
Sec.  53.450 will be met without reliance on human actions to achieve 
acceptable event mitigation.
    (3) The functional requirements analysis and function allocation 
performed under Sec.  53.730(d) must demonstrate that functions 
required for safety are not reliant upon credited human action.
    (4) The plant response to events analyzed under Sec.  53.450 must 
rely exclusively on safety features and characteristics that will 
neither be rendered unavailable by credible human errors of commission 
or omission nor credibly require manual human operation in response to 
equipment failures. Compliance with this paragraph may be achieved 
through the use of SSCs that function through inherent characteristics 
or that have engineered protections against human failures.
    (5) The plant design must provide for a layered defense-in-depth 
approach that is not dependent upon any single barrier or credited 
human action.
    (b) [Reserved]


Sec.  53.805  Facility licensee requirements related to generally 
licensed reactor operators.

    (a) Licensees for self-reliant-mitigation facilities that have not 
yet certified the permanent cessation of operations and permanent 
removal of fuel from the reactor vessel as described under Sec.  
53.1070 must demonstrate compliance with the following requirements:
    (1) Ensure that, in addition to being qualified to perform those 
items identified by the facility-specific systems approach to training 
conducted under Sec.  53.815, generally licensed reactor operators are 
qualified to safely and competently--
    (i) Perform administrative tasks, including compliance with 
technical specifications, and perform operability determinations;
    (ii) Implement maintenance and configuration controls;
    (iii) Comply with radioactive release limitations;
    (iv) Understand plant operating data, including reactor parameters, 
and evaluate emergency conditions;
    (v) Initiate a reactor shutdown from necessary locations;
    (vi) Dispatch and direct operations and maintenance personnel;
    (vii) Implement any applicable responsibilities under the facility 
emergency plan; and
    (viii) Make required notifications to local, State, participating 
Tribal and Federal authorities.
    (2) Develop, implement, and maintain facility technical 
specifications that provide the necessary administrative controls to 
ensure the implementation of these requirements.
    (3) Develop, implement, and maintain the generally licensed reactor 
operator training, examination, and proficiency programs required under 
Sec.  53.815.
    (4) Ensure that generally licensed reactor operators are subject to 
the facility's generally licensed reactor operator training, 
examination, and proficiency programs required under Sec.  53.815. 
Ensure that generally licensed reactor operators are subject to and 
comply with the applicable programmatic requirements for plant 
personnel required under 10 CFR parts 26 and 73. An individual that is 
not in compliance with any of these programs is not qualified to be in 
a position that may involve the manipulation of the controls of the 
commercial nuclear plant.
    (5) Report annually to the NRC the identity of all generally 
licensed reactor operators at the commercial nuclear plant, including 
all additions and deletions since the previous report.

[[Page 87068]]

    (6) Ensure that the facility design continues to meet the criteria 
of Sec.  53.800.
    (b) [Reserved]


Sec.  53.810  Generally licensed reactor operators.

    (a) A general license to manipulate the controls of a self-reliant-
mitigation facility and to direct the licensed activities of generally 
licensed reactor operators is hereby issued to any individual employed 
in a position that may involve the manipulation of the controls of that 
self-reliant-mitigation facility and who observes the restrictions of 
this section.
    (b) A generally licensed reactor operator must comply with the 
operating procedures and other conditions specified in the license 
authorizing operation of the facility.
    (c) The general license is limited to the facility or facilities at 
which the operator is employed.
    (d) The Commission will suspend the general license on an 
individual basis for violations of any provision of the Act or any rule 
or regulation issued thereunder whenever the Commission deems such 
suspension desirable, including--
    (1) For willful violation of, or failure to observe, any of the 
terms and conditions of the Act or the general license, or of any rule, 
regulation, or order of the Commission;
    (2) For any conduct determined by the Commission to be a hazard to 
safe operation of the facility; or
    (3) For the sale, use, or possession of illegal drugs, or refusal 
to participate in the facility drug and alcohol testing program, or a 
confirmed positive test for drugs, drug metabolites, or alcohol in 
violation of the conditions and cutoff levels established by Sec.  
53.810(f) or the consumption of alcoholic beverages within the 
protected area of commercial nuclear plants, or a determination of 
unfitness for scheduled work as a result of the consumption of 
alcoholic beverages.
    (e) The Commission may require information from a generally 
licensed reactor operator to determine whether a general license should 
be revoked or suspended with respect to that operator.
    (f) The generally licensed reactor operator must not consume or 
ingest alcoholic beverages within the protected area of commercial 
nuclear plants. The generally licensed reactor operator must not use, 
possess, or sell any illegal drugs. The generally licensed reactor 
operator must not perform activities requiring a general license while 
under the influence of alcohol or any prescription, over-the-counter, 
or illegal substance that could adversely affect his or her ability to 
safely and competently perform these activities. For the purpose of 
this paragraph, with respect to alcoholic beverages and drugs, the term 
``under the influence'' means the generally licensed reactor operator 
exceeded, as evidenced by a confirmed test result, the lower of the 
cutoff levels for drugs or alcohol contained in 10 CFR part 26, or as 
established by the facility licensee. The term ``under the influence'' 
also means the generally licensed reactor operator could be mentally or 
physically impaired as a result of substance use including prescription 
and over-the-counter drugs, as determined under the provisions, 
policies, and procedures established by the facility licensee for its 
fitness-for-duty program, in such a manner as to adversely affect his 
or her ability to safely and competently perform generally licensed 
reactor operator duties.
    (g) The generally licensed reactor operator must notify the 
Commission within 30 days about a conviction for a felony.


Sec.  53.815  Generally licensed reactor operator training, 
examination, and proficiency programs.

    (a) Applicability. The requirements of this section apply to each 
licensee of a self-reliant-mitigation facility that has not yet 
certified the permanent cessation of operations and permanent removal 
of fuel from the reactor vessel as described under Sec.  53.1070.
    (b) Requirements. (1) The licensee must develop, implement, and 
maintain training and examination programs that demonstrate compliance 
with the requirements of paragraphs (b)(2) and (3) of this section.
    (2) The training program must provide for both the initial and 
continuing training of generally licensed reactor operators and be 
derived from a systems approach to training as defined in this part.
    (3)(i) The training program must incorporate the instructional 
requirements necessary to provide qualified generally licensed reactor 
operators to operate and maintain the facility in a safe manner in all 
modes of operation. The training program must comply with the facility 
license, including all technical specifications and applicable 
regulations. The facility licensee must periodically evaluate and 
revise the training program as appropriate to reflect industry 
experience and relevant changes, including changes to the facility, 
procedures, regulations, and quality assurance (QA) requirements. 
Facility licensee management must periodically review the training 
program for effectiveness.
    (ii) The training program must ensure that generally licensed 
reactor operators have and maintain the necessary knowledge, skills, 
and abilities.
    (iii) The training program must include the generally licensed 
reactor operator manipulating the controls of either the facility or a 
simulation facility that demonstrates compliance with the requirements 
of Sec.  53.815(e).
    (iv) The training program must include an initial examination 
program for testing a representative sample of the knowledge, skills, 
and abilities needed to safely perform generally licensed reactor 
operator duties, to include both the examination methods and criteria 
to be used to assess passing performance. The facility licensee must 
provide the opportunity for a representative of the Commission to be 
present during initial examination administration.
    (v) The training program must include a requalification examination 
program for testing a sample of the topics included under the systems 
approach to training, to include the examination methods and criteria 
to be used to assess passing performance. The requalification 
examination program must specify an appropriate periodicity for 
administering a complete requalification examination to each generally 
licensed reactor operator, and the facility licensee must provide the 
opportunity for a representative of the Commission to be present during 
requalification examination administration.
    (A) The facility licensee must ensure that any generally licensed 
reactor operator who either demonstrates unsatisfactory performance on, 
or fails to complete, the requalification examination is removed from 
the performance of generally licensed reactor operator duties until 
such time that any necessary remedial training has been completed and a 
retake examination has been passed.
    (B) [Reserved]
    (vi) The training program must be approved by the Commission prior 
to its use, as described under Sec.  53.730(g). The examination program 
must provide for valid and reliable examinations and must be approved 
by the Commission prior to their use, as described under Sec.  
53.730(g). The approved programs are subject to the requirements of 
Sec.  53.1565.
    (c) Records. The following is required regarding the documentation 
of the generally licensed reactor operator training and examination 
programs:
    (1) Sufficient records must be maintained by the facility licensee 
to

[[Page 87069]]

maintain the integrity of the programs and kept available for NRC 
inspection to verify the adequacy of the programs.
    (2) The facility licensee must maintain records documenting the 
participation of each generally licensed reactor operator in the 
training and examination programs. The records must contain copies of 
examinations administered, the answers given by the generally licensed 
reactor operator, documentation of the grading of examinations, and 
documentation of any additional training administered in areas in which 
a generally licensed reactor operator exhibited deficiencies. The 
facility licensee must retain these records while the associated 
generally licensed reactor operators remain employed at the facility.
    (3) Each record required by this part must be legible throughout 
the retention period. The record may be the original, a reproduced 
copy, or an electronic copy provided that the copy is authenticated by 
authorized personnel.
    (d) Examination integrity. Generally licensed reactor operators and 
facility licensees must not engage in any activity that compromises the 
integrity of any examination conducted under the generally licensed 
reactor operator training and examination programs. The integrity of an 
examination is considered compromised if any activity, regardless of 
intent, affected, or, but for detection, could have affected the 
equitable and consistent administration of the examination. This 
includes all activities related to the preparation, administration, and 
grading of examinations.
    (e) Simulation facilities. (1) Simulation facilities used for 
training purposes, for maintaining proficiency, or for the conduct of 
examinations must demonstrate compliance with the following criteria as 
they relate to the facility licensee's reference plant:
    (i) The simulation facility must be of sufficient scope and 
fidelity for individuals to acquire and demonstrate the necessary 
knowledge, skills, and abilities to safely perform generally licensed 
reactor operator duties.
    (ii) The simulation facility must utilize models relating to 
nuclear, thermal-hydraulic, and other applicable design-specific 
characteristics that either replicate the most recent fuel load in the 
reference commercial nuclear plant or, prior to initial fuel load (or, 
for a fueled manufactured reactor, prior to initiating the physical 
removal of any one of the independent physical mechanisms to prevent 
criticality required under Sec.  53.620(d)(1)), replicate the intended 
initial fuel load for the reference commercial nuclear plant, with the 
exception of those portions of the simulation facility that utilize the 
reference plant itself.
    (iii) Simulator fidelity must be demonstrated so that significant 
control manipulations are completed without procedural exceptions, 
simulator performance exceptions, or deviation from the approved 
training scenario sequence.
    (2) Facility licensees that maintain a simulation facility for 
training purposes, for maintaining proficiency, or for the conduct of 
examinations must--
    (i) Conduct performance testing throughout the life of the 
simulation facility in a manner sufficient to ensure that paragraph 
(e)(1) of this section is met;
    (ii) Retain the results of performance testing for 4 years after 
the completion of each performance test or until superseded by updated 
test results;
    (iii) Promptly correct modeling and hardware discrepancies and 
discrepancies identified from scenario validation and from performance 
testing or provide justification for why the presence of such 
discrepancies will not adversely affect the criteria of paragraph 
(e)(1) of this section;
    (iv) Make the results of any uncorrected performance test failures 
that may exist at the time of an inspection available for NRC review; 
and
    (v) Maintain the provisions for examination integrity consistent 
with Sec.  53.815(d).
    (f) Waiver of examination requirement. The facility licensee may 
waive any or all of the requirements for an examination in accordance 
with the facility licensee's Commission-approved generally licensed 
reactor operator training and examination programs.
    (g) Proficiency. The facility licensee must develop, implement, and 
maintain a proficiency program to allow generally licensed reactor 
operators to maintain proficiency regarding position functions and 
familiarity with plant status. This program must include those steps 
that will be taken in order to re-establish proficiency when it cannot 
be maintained.


Sec.  53.820  Cessation of individual applicability.

    The general license ceases to be applicable on an individual basis 
once a generally licensed reactor operator is no longer being employed 
in a position that may involve the manipulation of the controls of the 
self-reliant mitigation facility.


Sec.  53.830  Training and qualification of commercial nuclear plant 
personnel.

    (a) This section addresses personnel training requirements. The 
regulations within this section are applicable to all applicants for or 
holders of OLs or COLs under this part.
    (b) Prior to initial fuel load (or, for a fueled manufactured 
reactor, prior to initiating the physical removal of any one of the 
independent physical mechanisms to prevent criticality required under 
Sec.  53.620(d)(1)), each holder of an operating or COL under this part 
must, with sufficient time to provide trained and qualified personnel 
to operate the facility, establish, implement, and maintain a training 
program that demonstrates compliance with the requirements of 
paragraphs (c) and (d) of this section.
    (c) The training program must be derived from a systems approach to 
training as defined in this part and must provide, at a minimum, for 
the training and qualification of the following categories of 
commercial nuclear plant personnel:
    (1) Supervisors (e.g., shift supervisors);
    (2) Technicians (e.g., maintenance, chemistry, and radiological); 
and
    (3) Other appropriate operating personnel (e.g., auxiliary 
operators, certified fuel handlers, and individuals who provide 
engineering expertise to on-shift operating personnel).
    (d) The training program must incorporate the instructional 
requirements necessary to provide qualified personnel to operate 
components of a commercial nuclear plant and maintain the facility in a 
safe manner in all modes of operation. The training program must be 
developed to be in compliance with the facility license, including all 
technical specifications and applicable regulations.
    (1) The training program must be periodically evaluated and revised 
as appropriate to reflect industry experience and relevant changes, 
including changes to the facility, procedures, regulations, and QA 
requirements. The training program must be periodically reviewed by 
facility licensee management for effectiveness.
    (2) Sufficient records must be maintained by the facility licensee 
to maintain program integrity and kept available for NRC inspection to 
verify the adequacy of the training program.


Sec.  53.845  Programs.

    (a) The required plant programs under this part must include but 
are not necessarily limited to the programs

[[Page 87070]]

described in the following sections of this subpart. Licensees may 
combine, separate, and otherwise organize programs and related 
documents as appropriate for the technologies and organizations 
associated with the commercial nuclear plant.
    (b) In addition to the programs described in the following 
sections, programs must be provided for each commercial nuclear plant, 
if necessary, to ensure that the performance of design features and 
human actions are consistent with the analyses performed under 
Sec. Sec.  53.450 and 53.730 and that the plant will demonstrate 
compliance with the safety criteria defined in Sec. Sec.  53.210 and 
53.220.


Sec.  53.850  Radiation protection.

    (a) Each holder of an OL or COL under this part must develop, 
implement, and maintain a Radiation Protection Program for operations 
that is commensurate with the scope and extent of licensed activities 
under this part and includes measures for limiting and monitoring 
radioactive plant effluents and limiting and monitoring the dose to 
individuals working with radioactive materials in accordance with 10 
CFR part 20.
    (b) Each holder of an OL or COL under this part must develop, 
implement, and maintain a program for the control of radioactive 
effluents and for keeping the doses to members of the public from 
radioactive effluents as low as is reasonably achievable and for 
environmental monitoring. The program must be contained in an Offsite 
Dose Calculations Manual, must be implemented by procedures, and must 
include remedial actions to be taken whenever the program limits are 
exceeded. The Offsite Dose Calculations Manual must--
    (1) Contain the methodology and parameters used in the calculation 
of offsite doses resulting from radioactive gaseous and liquid 
effluents, in the calculation of gaseous and liquid effluent monitoring 
alarm and trip setpoints, and in the conduct of the radiological 
environmental monitoring program; and
    (2) Contain the radioactive effluent controls and radiological 
environmental monitoring activities, and descriptions of the 
information that should be included in the Annual Radiological 
Environmental Operating and Radioactive Effluent Release Reports 
required by Sec.  53.1645.
    (c) Each holder of an OL or COL under this part must develop, 
implement, and maintain a Process Control Program that identifies the 
administrative and operational controls for solid radioactive waste 
processing, process parameters, and surveillance requirements 
sufficient to ensure compliance with the requirements of 10 CFR part 
20, 10 CFR part 61, and 10 CFR part 71.


Sec.  53.855  Emergency preparedness.

    (a) Each holder of an OL or COL under this part must have an 
emergency response plan that must contain information needed to 
demonstrate compliance with either the requirements in Sec.  50.160 of 
this chapter or the requirements in appendix E to part 50 and the 
planning standards of Sec.  50.47(b) of this chapter.
    (b) No initial OL, initial COL, or early site permit that includes 
complete and integrated emergency plans will be issued under this part 
unless a finding is made by the NRC, in accordance with Sec.  50.47 of 
this chapter, that there is reasonable assurance that adequate 
protective measures can and will be taken in the event of a 
radiological emergency.


Sec.  53.860  Security programs.

    (a) Physical Protection Program. Each holder of an OL or COL under 
this part must develop, implement, and maintain a physical protection 
program under the following requirements:
    (1) The licensee must implement security requirements for the 
protection of special nuclear material based on the type, enrichment, 
and quantity in accordance with 10 CFR part 73, as applicable, and 
implement security requirements for the protection of Category 1 and 
Category 2 quantities of radioactive material in accordance with 10 CFR 
part 37, as applicable; and
    (2) The licensee must demonstrate compliance with the provisions 
set forth in either Sec. Sec.  73.55 or 73.100 of this chapter, unless 
the licensee demonstrates compliance with the following criterion:
    (i) The radiological consequences from a design-basis threat-
initiated event involving the loss of engineered systems for decay heat 
removal and possible breaches in physical structures surrounding the 
reactor, spent fuel, and other inventories of radioactive materials 
result in offsite doses below the values in Sec.  53.210.
    (ii) The applicant must perform a site-specific analysis, including 
identification of target sets, to demonstrate that the criterion in 
Sec.  53.860(a)(2)(i) is satisfied. The analysis must assume that 
licensee mitigation and recovery actions, including any operator 
actions, are unavailable or ineffective. The licensee must maintain the 
analysis until the permanent cessation of operations and permanent 
removal of fuel from the reactor vessel as described under Sec.  
53.1070.
    (b) Fitness for Duty. Each holder of an OL or COL under this part 
must develop, implement, and maintain a fitness for duty program under 
10 CFR part 26.
    (c) Access Authorization. Each holder of an OL or COL under this 
part must develop, implement, and maintain an access authorization 
program under Sec.  73.120 of this chapter if the criterion in Sec.  
53.860(a)(2)(i) is satisfied, or the requirements in Sec.  73.56 of 
this chapter if the criterion is not satisfied.
    (d) Cybersecurity. Each holder of an OL or COL under this part must 
develop, implement, and maintain a cybersecurity program under 
Sec. Sec.  73.54 or 73.110 of this chapter.
    (e) Information Security. Each holder of an OL or COL under this 
part must develop, implement, and maintain an information protection 
system under Sec. Sec.  73.21, 73.22, and 73.23 of this chapter, as 
applicable.


Sec.  53.865  Quality assurance.

    Each holder of an OL or COL under this part must develop, 
implement, and maintain a quality assurance program in accordance with 
appendix B of part 50 of this chapter. A written quality assurance 
program manual must be developed and used to guide the conduct of the 
program in accordance with generally accepted consensus codes and 
standards that have been endorsed or otherwise found acceptable by the 
NRC.


Sec.  53.870  Integrity assessment programs.

    Each holder of an OL or COL under this part must develop, 
implement, and maintain an integrity assessment program to monitor, 
evaluate, and manage--
    (a) The effects of plant aging on SR and NSRSS SSCs. The program 
may refer to surveillances, tests, and inspections conducted for 
specific SSCs in accordance with other requirements in this part or 
conducted in accordance with applicable consensus codes and standards 
endorsed or otherwise found acceptable by the NRC;
    (b) Cyclic or transient load limits to ensure that SR and NSRSS 
SSCs are maintained within the applicable design limits; and
    (c) Degradation mechanisms related to chemical interactions, 
operating temperatures, effects of irradiation, and other environmental 
factors to ensure that the capabilities, availability, and reliability 
of SR and NSRSS SSCs demonstrate compliance with the

[[Page 87071]]

functional design criteria of Sec. Sec.  53.410 and 53.420.


Sec.  53.875  Fire protection.

    (a)(1) Each holder of an OL or COL under this part must have a fire 
protection plan that describes the overall fire protection program for 
the facility; identifies the various positions within the licensee's 
organization that are responsible for the program; states the 
authorities that are delegated to each of these positions to implement 
those responsibilities; and outlines the plans for fire protection, 
fire detection and suppression capability; and limitation of fire 
damage.
    (2) The fire protection plan must also describe specific features 
necessary to implement the program described in paragraph (a)(1) of 
this section such as the following: administrative controls and 
personnel requirements for fire prevention and manual fire suppression 
activities; automatic and manually operated fire detection and 
suppression systems; and the means to limit fire damage to SSCs so that 
the capability to demonstrate compliance with the requirements of Sec.  
53.210 is ensured.
    (b)(1) Each holder of an OL or COL under this part must develop a 
performance-based or deterministic fire protection program that 
demonstrates compliance with the safety criteria outlined in Sec. Sec.  
53.210 and 53.220, related safety functions outlined in Sec.  53.230, 
and defense in depth as outlined in Sec.  53.250 with specific fire 
protection measures related to fire prevention, fire detection, and 
fire suppression.
    (2) The fire protection program must comply with the following:
    (i) Safety-related and NSRSS SSCs must be designed, located, and 
maintained to minimize, consistent with other safety requirements, the 
probability and effect of fires and explosions.
    (ii) Noncombustible and fire-resistant materials must be used 
wherever practical throughout the facility, particularly in locations 
with SR and NSRSS SSCs.
    (iii) Fire detection and fire suppression systems of appropriate 
capacity and capability must be provided and designed and maintained to 
minimize the adverse effects of fires on SR and NSRSS SSCs.
    (iv) Fire suppression systems must be designed and maintained to 
ensure that their rupture or inadvertent operation does not 
significantly impair the ability of SR and NSRSS SSCs to perform their 
safety functions to satisfy Sec.  53.230.


Sec.  53.880  Inservice inspection and inservice testing.

    (a) Each holder of an OL or COL under this part must develop, 
implement, and maintain a program for inservice inspection (ISI) and 
inservice testing (IST) prior to receiving an OL or COL. The ISI/IST 
programs must, wherever applicable, be in accordance with generally 
accepted consensus codes and standards that have been endorsed or 
otherwise found acceptable by the NRC. The ISI/IST program must include 
all inspections and tests required by the codes and standards used in 
the design and be supplemented by risk insights that identify the most 
important SSCs to plant safety. The types of testing and inspections 
and their frequency should be informed by risk insights to maintain the 
reliability and performance of SSCs consistent with the associated 
design and analyses activities involving those SSCs. Risk insights must 
also be used to determine when to conduct the inspections and tests 
(e.g., full power, shutdown, refueling) to minimize risk to the plant 
workers and the public. The ISI/IST program must be documented in a 
written manual and managed by qualified personnel reporting to the 
Plant Manager.
    (b) Prior to plant operation, baseline inspections and testing must 
be performed using the same techniques as will be used for future 
inspections and testing. The results of these inspections and testing 
must be used as benchmarks for evaluating the results of future 
inspections and testing. Sufficient room and support must be provided 
to accommodate the personnel, ISI/IST equipment, and shielding 
necessary to perform the inspections and testing. Acceptance criteria 
for determining whether corrective action is needed must be developed 
(or taken from the codes and standards used in the design) for 
evaluating the results of the inspections and testing. The results of 
the inspections and testing must be provided to the Plant Manager who 
is responsible for determining what, if any, corrective action is 
needed and when it should be taken. The ISI/IST results and corrective 
actions must be documented and the documentation retained for the life 
of the plant.


Sec.  53.910  Procedures and guidelines.

    (a) Each holder of an OL or COL under this part must have a program 
for developing, implementing, and maintaining an integrated set of 
procedures, guidelines, and related supporting activities to support 
normal operations and respond to possible unplanned events.
    (b) The program required by paragraph (a) of this section must 
include but is not limited to development, implementation, maintenance, 
and supporting activities of procedures and guidelines for the 
following:
    (1) Plant operations;
    (2) Maintenance activities under Sec.  53.715;
    (3) Program requirements under this subpart F of this part;
    (4) Emergency operating procedures, if developed to address the 
role of human actions in responding to LBEs;
    (5) Accident management guidelines, if developed to address the 
role of human actions in responding to LBEs;
    (6) Procedures for each area in which licensed special nuclear 
material is handled, used, or stored to protect personnel upon the 
sounding of a criticality alarm required by Sec.  53.440(m); and
    (7) Procedures that describe how the licensee will address the 
following areas if the licensee is notified of a potential aircraft 
threat:
    (i) Verification of the authenticity of threat notifications;
    (ii) Maintenance of continuous communication with threat 
notification sources;
    (iii) Contacting all onsite personnel and applicable offsite 
response organizations;
    (iv) Onsite actions necessary to enhance the capability of the 
facility to mitigate the consequences of an aircraft impact;
    (v) Measures to reduce visual discrimination of the site relative 
to its surroundings or individual buildings within the protected area;
    (vi) Dispersal of equipment and personnel, as well as rapid entry 
into site protected areas for essential onsite personnel and offsite 
responders who are necessary to mitigate the event; and
    (vii) Recall of site personnel.

Subpart G--Decommissioning Requirements


Sec.  53.1000  Scope and purpose.

    This subpart defines the requirements related to decommissioning 
for applicants for, or holders of, an operating license (OL) or 
combined license (COL). The requirements related to maintaining 
financial assurance for decommissioning are in Sec. Sec.  53.1010 
through 53.1060. The requirements for transitioning from operations to 
decommissioning and for the release of property and termination of the 
license are in Sec. Sec.  53.1070 through 53.1080.

[[Page 87072]]

Sec.  53.1010  Financial assurance for decommissioning.

    (a) This section establishes requirements for indicating to the 
U.S. Nuclear Regulatory Commission (NRC) how an applicant for or holder 
of an OL or COL under this part will provide reasonable assurance that 
funds will be available for the decommissioning process. Reasonable 
assurance consists of a series of steps as provided in paragraph (b) of 
this section and Sec. Sec.  53.1020, 53.1030 and 53.1040. Funding for 
the decommissioning of commercial nuclear plants may also be subject to 
the regulation of Federal or State government agencies (e.g., Federal 
Energy Regulatory Commission (FERC) and State Public Utility 
Commissions) that have jurisdiction over rate regulation. The 
requirements of this subpart, in particular Sec.  53.1020, are in 
addition to, and not a substitution for, other requirements, and are 
not intended to be used by themselves or by other agencies to establish 
rates.
    (b) Each applicant for an OL or COL under this part must prepare a 
plan and an associated decommissioning report that ensures and 
documents that adequate funding will be available to decommission the 
facility. Each holder of an OL or COL must implement and maintain the 
plan.
    (1)(i) Before the Commission issues an OL under this part, the 
applicant must update the decommissioning report to certify that it has 
provided financial assurance for decommissioning in the amount proposed 
in the application and approved by the NRC under Sec.  53.1020.
    (ii) No later than 30 days after the Commission issues the notice 
of intended operation under Sec.  53.1452 for a COL under this part, 
the licensee must update the decommissioning report to certify that it 
has provided financial assurance for decommissioning in the amount 
proposed in the application and approved by the NRC under Sec.  
53.1020.
    (2) The amount of financial assurance for decommissioning to be 
provided must be based on a site-specific cost estimate for 
decommissioning the facility under Sec.  53.1020.
    (3) The amount of financial assurance for decommissioning to be 
provided must be adjusted annually using a rate at least equal to that 
stated in Sec.  53.1030.
    (4) The amount of financial assurance for decommissioning to be 
provided must be covered by one or more of the methods described in 
Sec.  53.1040 as acceptable to the NRC. A copy of the financial 
instrument obtained to satisfy the requirements of Sec.  53.1040 must 
be submitted to the NRC as part of the application for an OL under this 
part; however, an applicant for or holder of a COL need not obtain such 
financial instrument or submit a copy to the Commission except as 
provided in Sec.  53.1060(b).


Sec.  53.1020  Cost estimates for decommissioning.

    Cost estimates for decommissioning must be site-specific. Site-
specific decommissioning cost estimates (DCEs) must account for the 
engineering, labor, equipment, transportation, disposal, and related 
charges needed to support termination of the license. They must include 
the costs for decontaminating structures, systems, and components and 
the site environs; removal of contaminated components and materials 
from the plant and the site environs; disposal of removed components 
and materials in appropriate facilities; and any other activities 
supporting the release of the property and termination of the license. 
They must also address the approach to annual adjustments required by 
Sec.  53.1030. Finally, site-specific DCEs must include plans for 
adjusting levels of funds assured for decommissioning to demonstrate 
that a reasonable level of assurance will be provided that funds will 
be available when needed to cover the cost of decommissioning.


Sec.  53.1030  Annual adjustments to cost estimates for 
decommissioning.

    Each holder of an OL or COL under this part must annually adjust 
the cost estimate for decommissioning to account for escalation in 
labor, energy, and waste burial costs. Licensees may elect to use 
either a site-specific adjustment factor, approved as part of the plan 
and associated decommissioning report required by Sec.  53.1010, in 
paragraph (a) of this section or the generic adjustment factor in 
paragraph (b) of this section.
    (a) A site-specific adjustment factor must address the estimated 
contributions and escalation of costs for the following aspects of 
decommissioning:
    (1) Labor, materials, and services;
    (2) Energy and waste transportation; and
    (3) Radioactive waste burial or other disposition.
    (b) A generic adjustment factor must be at least equal to 0.65 L + 
0.13 E + 0.22 B, where L and E are escalation factors for labor and 
energy, respectively, and are to be taken from regional data of U.S. 
Department of Labor Bureau of Labor Statistics and B is an escalation 
factor for waste burial and is to be taken from NRC report NUREG-1307, 
``Report on Waste Burial Charges.''


Sec.  53.1040  Methods for providing financial assurance for 
decommissioning.

    Financial assurance for decommissioning is to be provided by the 
following methods.
    (a) Prepayment. Prepayment is the deposit made preceding the start 
of operation or the transfer of a license under Sec.  53.1570 into an 
account segregated from licensee assets and outside the administrative 
control of the licensee and its subsidiaries or affiliates of cash or 
liquid assets such that the amount of funds would be sufficient to pay 
decommissioning costs. Prepayment may be in the form of a trust, escrow 
account, or Government fund with payment by certificate of deposit, 
deposit of government or other securities, or other method acceptable 
to the NRC. This trust, escrow account, Government fund, or other type 
of agreement must be established in writing and maintained at all times 
in the United States with an entity that is an appropriate State or 
Federal Government agency, or an entity whose operations in which the 
prepayment deposit is managed are regulated and examined by a Federal 
or State agency. A licensee that has prepaid funds based on a site-
specific cost estimate under Sec.  53.1020 may take credit for 
projected earnings on the prepaid decommissioning trust funds, using up 
to a 2 percent annual real rate of return through the time of 
termination of the license. A licensee may use a credit of greater than 
2 percent if the licensee's rate-setting authority has specifically 
authorized a higher rate. Actual earnings on existing funds may be used 
to calculate future fund needs.
    (b) External sinking fund. An external sinking fund is a fund 
established and maintained by setting funds aside periodically in an 
account segregated from licensee assets and outside the administrative 
control of the licensee and its subsidiaries or affiliates in which the 
total amount of funds would be sufficient to pay decommissioning costs. 
An external sinking fund may be in the form of a trust, escrow account, 
or Government fund, with payment by certificate of deposit, deposit of 
Government or other securities, or other method acceptable to the NRC. 
This trust, escrow account, Government fund, or other type of agreement 
must be established in writing and maintained at all times in the 
United States with an entity that is an appropriate State or Federal 
Government agency, or an entity whose operations in which the external 
sinking fund is managed are regulated and examined by a Federal or 
State agency. A licensee that has collected funds based on a site-
specific cost

[[Page 87073]]

estimate under Sec.  53.1020 may take credit for projected earnings on 
the external sinking funds using up to a 2 percent annual real rate of 
return from the time of future funds' collection through the time of 
termination of the license. A licensee may use a credit of greater than 
2 percent if the licensee's rate-setting authority has specifically 
authorized a higher rate. Actual earnings on existing funds may be used 
to calculate future fund needs. A licensee whose rates for 
decommissioning costs cover only a portion of these costs may make use 
of this method only for the portion of these costs that are collected 
in one of the manners described in this paragraph. This method may be 
used as the exclusive mechanism relied upon for providing financial 
assurance for decommissioning in the following circumstances:
    (1) By a licensee that recovers, either directly or indirectly, the 
estimated total cost of decommissioning through rates established by 
``cost of service'' or similar ratemaking regulation. Public utility 
districts, municipalities, rural electric cooperatives, and State and 
Federal agencies, including associations of any of the foregoing, that 
establish their own rates and are able to recover their cost of service 
allocable to decommissioning, are deemed to satisfy this condition.
    (2) By a licensee whose source of revenues for its external sinking 
fund is a ``non-bypassable charge,'' the total amount of which will 
provide funds estimated to be needed for decommissioning pursuant to 
Sec. Sec.  53.1020, 53.1060, or 53.1575.
    (c) A surety method, insurance, or other guarantee method. (1) 
These methods guarantee that decommissioning costs will be paid. A 
surety method may be in the form of a surety bond, or letter of credit. 
Any surety method or insurance used to provide financial assurance for 
decommissioning must contain the following conditions:
    (i) The surety method or insurance must be open-ended, or, if 
written for a specified term, such as 5 years, must be renewed 
automatically, unless 90 days or more prior to the renewal day the 
issuer notifies the NRC, the beneficiary, and the licensee of its 
intention not to renew. The surety or insurance must also provide that 
the full-face amount be paid to the beneficiary automatically prior to 
the expiration without proof of forfeiture if the licensee fails to 
provide a replacement acceptable to the NRC within 30 days after 
receipt of notification of cancellation.
    (ii) The surety or insurance must be payable to a trust established 
for decommissioning costs. The trustee and trust must be acceptable to 
the NRC. An acceptable trustee includes an appropriate State or Federal 
Government agency or an entity that has the authority to act as a 
trustee and whose trust operations are regulated and examined by a 
Federal or State agency.
    (2) A parent company guarantee of funds for decommissioning costs 
based on a financial test may be used if the guarantee and test are as 
contained in appendix A to 10 CFR part 30.
    (3) For commercial companies that issue bonds, a guarantee of funds 
by the applicant or licensee for decommissioning costs based on a 
financial test may be used if the guarantee and test are as contained 
in appendix C to 10 CFR part 30. For commercial companies that do not 
issue bonds, a guarantee of funds by the applicant or licensee for 
decommissioning costs may be used if the guarantee and test are as 
contained in appendix D to 10 CFR part 30. A guarantee by the applicant 
or licensee may not be used in any situation in which the applicant or 
licensee has a parent company holding majority control of voting stock 
of the company.
    (d) Funding method for Federal licensees. For a Federal licensee, a 
statement of intent containing a cost estimate for decommissioning and 
indicating that funds for decommissioning will be obtained when 
necessary.
    (e) Contractual funding method. Contractual obligation(s) on the 
part of a licensee's customer(s), the total amount of which over the 
duration of the contract(s) will provide the licensee's total share of 
uncollected funds estimated to be needed for decommissioning pursuant 
to Sec. Sec.  53.1020, 53.1060, or 53.1575. To be acceptable to the NRC 
as a method of decommissioning funding assurance, the terms of the 
contract(s) must include provisions that the buyer(s) of electricity or 
other products will pay for the decommissioning obligations specified 
in the contract(s), notwithstanding the operational status either of 
the licensed plant to which the contract(s) pertains or force majeure 
provisions. All proceeds from the contract(s) for decommissioning 
funding will be deposited to the external sinking fund. The NRC 
reserves the right to evaluate the terms of any contract(s) and the 
financial qualifications of the contracting entity or entities offered 
as assurance for decommissioning funding.
    (f) Other funding mechanisms. Any other mechanism, or combination 
of mechanisms, that provides, as determined by the NRC upon its 
evaluation of the specific circumstances of each licensee submittal, 
assurance of decommissioning funding equivalent to that provided by the 
mechanisms specified in paragraphs (a) through (e) of this section. 
Licensees who do not have sources of funding described in paragraph (b) 
of this section may use an external sinking fund in combination with a 
guarantee mechanism, as specified in paragraph (c) of this section, 
provided that the total amount of funds estimated to be necessary for 
decommissioning is assured.


Sec.  53.1045  Limitations on the use of decommissioning trust funds.

    (a)(1) Decommissioning trust funds may be used by licensees if--
    (i) The withdrawals are for expenses for decommissioning activities 
consistent with the definition of decommission or decommissioning in 
Sec.  53.020;
    (ii) The expenditure would not reduce the value of the 
decommissioning trust below an amount necessary to place and maintain 
the reactor in a safe storage condition if unforeseen conditions or 
expenses arise; and
    (iii) The withdrawals would not inhibit the ability of the licensee 
to complete funding of any shortfalls in the decommissioning trust 
needed to ensure the availability of funds to ultimately release the 
site and terminate the license.
    (2) Initially, 3 percent of the amount determined in accordance 
with Sec.  53.1020 may be used for decommissioning planning. For 
licensees that have submitted the certifications required under Sec.  
53.1070 and commencing 90 days after the NRC has received the post-
shutdown decommissioning activities report (PSDAR) required by Sec.  
53.1060, an additional 20 percent may be used. An updated site-specific 
DCE must be submitted to the NRC prior to the licensee using any 
funding in excess of these amounts.
    (b) Licensees that are not ``electric utilities'' as defined in 
Sec.  53.020 that use prepayment or an external sinking fund to provide 
financial assurance must provide in the terms of the arrangements 
governing the trust, escrow account, or Government fund, used to 
segregate and manage the funds that--
    (1) The trustee, manager, investment advisor, or other person 
directing investment of the funds--
    (i) Is prohibited from investing the funds in securities or other 
obligations of the licensee or any other owner or operator of any 
commercial nuclear

[[Page 87074]]

plant or their affiliates, subsidiaries, successors or assigns, or in a 
mutual fund in which at least 50 percent of the fund is invested in the 
securities of a licensee or parent company whose subsidiary is an owner 
or operator of a foreign or domestic commercial nuclear plant. However, 
the funds may be invested in securities tied to market indices or other 
non-nuclear sector collective, commingled, or mutual funds, provided 
that no more than 10 percent of trust assets may be indirectly invested 
in securities of any entity owning or operating one or more commercial 
nuclear plants.
    (ii) Is obligated at all times to adhere to a standard of care set 
forth in the trust, which either shall be the standard of care, whether 
in investing or otherwise, required by State or Federal law or one or 
more State or Federal regulatory agencies with jurisdiction over the 
trust funds, or, in the absence of any such standard of care, whether 
in investing or otherwise, that a prudent investor would use in the 
same circumstances. The term ``prudent investor,'' shall have the same 
meaning as set forth in FERC's ``Regulations Governing Nuclear Plant 
Decommissioning Trust Funds'' at 18 CFR 35.32(a)(3), or any successor 
regulation.
    (2) The licensee, its affiliates, and its subsidiaries are 
prohibited from being engaged as investment manager for the funds or 
from giving day-to-day management direction of the funds' investments 
or direction on individual investments by the funds, except in the case 
of passive fund management of trust funds where management is limited 
to investments tracking market indices.
    (3) The trust, escrow account, Government fund, or other account 
used to segregate and manage the funds may not be amended in any 
material respect without written notification to the Director, Office 
of Nuclear Reactor Regulation, or Director, Office of Nuclear Material 
Safety and Safeguards, as applicable, at least 30 working days before 
the proposed effective date of the amendment. The licensee must provide 
the text of the proposed amendment and a statement of the reason for 
the proposed amendment. The trust, escrow account, Government fund, or 
other account may not be amended if the person responsible for managing 
the trust, escrow account, Government fund, or other account receives 
written notice of objection from the Director, Office of Nuclear 
Reactor Regulation, or Director, Office of Nuclear Material Safety and 
Safeguards, as applicable, within the notice period.
    (4) Except for withdrawals being made under paragraph (a) of this 
section or for payments of ordinary administrative costs (including 
taxes) and other incidental expenses of the fund (including legal, 
accounting, actuarial, and trustee expenses) in connection with the 
operation of the fund, no disbursement or payment may be made from the 
trust, escrow account, Government fund, or other account used to 
segregate and manage the funds until written notice of the intention to 
make a disbursement or payment has been given to the Director, Office 
of Nuclear Reactor Regulation, or Director, Office of Nuclear Material 
Safety and Safeguards, as applicable, at least 30 working days before 
the date of the intended disbursement or payment. The disbursement or 
payment from the trust, escrow account, Government fund or other 
account may be made following the 30 working day notice period if the 
person responsible for managing the trust, escrow account, Government 
fund, or other account does not receive written notice of objection 
from the Director, Office of Nuclear Reactor Regulation, or Director, 
Office of Nuclear Material Safety and Safeguards, as applicable, within 
the notice period. Disbursements or payments from the trust, escrow 
account, Government fund, or other account used to segregate and manage 
the funds, other than for payment of ordinary administrative costs 
(including taxes) and other incidental expenses of the fund (including 
legal, accounting, actuarial, and trustee expenses) in connection with 
the operation of the fund, are restricted to decommissioning expenses 
or transfer to another financial assurance method acceptable under 
Sec.  53.1040 until final decommissioning has been completed. After 
decommissioning has begun and withdrawals from the decommissioning fund 
are made under paragraph (a) of this section, no further notification 
need be made to the NRC.
    (c) Licensees that are ``electric utilities'' under Sec.  53.020 
that use prepayment or an external sinking fund to provide financial 
assurance must include a provision in the terms of the trust, escrow 
account, Government fund, or other account used to segregate and manage 
funds that except for withdrawals being made under paragraph (a) of 
this section or for payments of ordinary administrative costs 
(including taxes) and other incidental expenses of the fund (including 
legal, accounting, actuarial, and trustee expenses) in connection with 
the operation of the fund, no disbursement or payment may be made from 
the trust, escrow account, Government fund, or other account used to 
segregate and manage the funds until written notice of the intention to 
make a disbursement or payment has been given the Director, Office of 
Nuclear Reactor Regulation, or Director, Office of Nuclear Material 
Safety and Safeguards, as applicable, at least 30 working days before 
the date of the intended disbursement or payment. The disbursement or 
payment from the trust, escrow account, Government fund or other 
account may be made following the 30 working day notice period if the 
person responsible for managing the trust, escrow account, Government 
fund, or other account does not receive written notice of objection 
from the Director, Office of Nuclear Reactor Regulation, or Director, 
Office of Nuclear Material Safety and Safeguards, as applicable, within 
the notice period. Disbursements or payments from the trust, escrow 
account, Government fund, or other account used to segregate and manage 
the funds, other than for payment of ordinary administrative costs 
(including taxes) and other incidental expenses of the fund (including 
legal, accounting, actuarial, and trustee expenses) in connection with 
the operation of the fund, are restricted to decommissioning expenses 
or transfer to another financial assurance method acceptable under 
Sec.  53.1040 until final decommissioning has been completed. After 
decommissioning has begun and withdrawals from the decommissioning fund 
are made under paragraph (a) of this section, no further notification 
need be made to the NRC.
    (d) A licensee that is not an ``electric utility'' under Sec.  
53.020 and using a surety method, insurance, or other guarantee method 
to provide financial assurance must provide that the trust established 
for decommissioning costs to which the surety or insurance is payable 
contains in its terms the requirements in Sec.  53.1045(b)(1) through 
(4).


Sec.  53.1050  NRC oversight.

    The NRC reserves the right to take the following steps in order to 
ensure a licensee's adequate accumulation of decommissioning funds: 
review, as needed, the rate of accumulation of decommissioning funds 
and, either independently or in cooperation with FERC and the 
licensee's State Public Utility Commission, take additional actions as 
appropriate on a case-by-case basis, including modification of a 
licensee's schedule for the accumulation of decommissioning funds.

[[Page 87075]]

Sec.  53.1060  Reporting and recordkeeping requirements.

    (a) Each holder of an OL under this part or holder of a COL under 
this part after the date that the Commission has made the finding under 
Sec.  53.1452(g) must report, at least once every 2 years, by March 31, 
on the status of its certification of decommissioning funding for each 
commercial nuclear reactor or part of a commercial nuclear reactor that 
it owns. The information in this report must include, at a minimum, the 
amount of decommissioning funds estimated to be required under 
Sec. Sec.  53.1020 and 53.1030; the amount of decommissioning funds 
accumulated to the end of the calendar year preceding the date of the 
report; a schedule of the annual amounts remaining to be collected; the 
assumptions used regarding rates of escalation in decommissioning 
costs, rates of earnings on decommissioning funds, and rates of other 
factors used in funding projections; any contracts upon which the 
licensee is relying under Sec.  53.1040(e); any modifications occurring 
to a licensee's method of providing financial assurance since the last 
submitted report; and any material changes to trust agreements. If any 
of the preceding items is not applicable, the licensee should so state 
in its report. Any licensee for a plant that is within 5 years of the 
projected end of its operation, or where conditions have changed such 
that it will close within 5 years (before the end of its licensed 
life), or that has already closed (before the end of its licensed 
life), or that is involved in a merger or an acquisition must submit 
this report annually.
    (b) Each holder of a COL under this part must, 2 years before and 1 
year before the scheduled date for initial loading of fuel (or, for a 
fueled manufactured reactor, 2 years before and 1 year before the 
scheduled date for initiating the physical removal of any one of the 
independent physical mechanisms to prevent criticality required under 
Sec.  53.620(d)(1)), submit a report to the NRC containing a 
certification updating the DCEs and a copy of the financial instrument 
to be used to satisfy Sec.  53.1040. No later than 30 days after the 
Commission publishes notice in the Federal Register under Sec.  
53.1452(a), the licensee must submit an updated decommissioning report 
required under Sec.  53.1010(b)(1)(ii), including a copy of the 
financial instrument obtained to satisfy Sec.  53.1040.
    (c) Each licensee must keep records of information important to the 
safe and effective decommissioning of the facility in an identified 
location until the license is terminated by the Commission. If records 
of relevant information are kept for other purposes, reference to these 
records and their locations may be used. Information the Commission 
considers important to decommissioning consists of--
    (1) Records of spills or other unusual occurrences involving the 
spread of contamination in and around the facility, equipment, or site. 
These records may be limited to instances when significant 
contamination remains after any cleanup procedures or when there is 
reasonable likelihood that contaminants may have spread to inaccessible 
areas as in the case of possible seepage into porous materials such as 
concrete. These records must include any known information on 
identification of involved nuclides, quantities, forms, and 
concentrations.
    (2) As-built drawings and modifications of structures and equipment 
in restricted areas where radioactive materials are used and/or stored 
and of locations of possible inaccessible contamination such as buried 
pipes that may be subject to contamination. If required drawings are 
referenced, each relevant document need not be indexed individually. If 
drawings are not available, the licensee must substitute appropriate 
records of available information concerning these areas and locations.
    (3) Records of the cost estimate performed for the decommissioning 
funding plan or of the amount certified for decommissioning, and 
records of the funding method used for assuring funds if either a 
funding plan or certification is used.
    (4) Records of--
    (i) The licensed site area, as originally licensed and any 
revisions, which must include a site map and any acquisition or use of 
property outside the originally licensed site area for the purpose of 
receiving, possessing, or using licensed materials;
    (ii) The licensed activities carried out on the acquired or used 
property; and
    (iii) The release and final disposition of any property recorded in 
paragraph (c)(4)(i) of this section, the historical site assessment 
performed for the release, radiation surveys performed to support 
release of the property, submittals to the NRC made under Sec.  
53.1070, and the methods employed to ensure that the property met the 
radiological criteria of subpart E of 10 CFR part 20 at the time the 
property was released.
    (d) Each holder of an OL or COL under this part must at or about 5 
years prior to the projected end of operations submit a preliminary DCE 
which includes an up-to-date assessment of the major factors that could 
affect the cost to decommission.
    (e) Prior to or within 2 years following permanent cessation of 
operations, the licensee must submit a PSDAR to the NRC, and a copy to 
the affected State(s). The PSDAR must contain a description of the 
planned decommissioning activities along with a schedule for their 
accomplishment, a discussion that provides the reasons for concluding 
that the environmental impacts associated with site-specific 
decommissioning activities will be bounded by appropriate previously 
issued environmental impact statements, and a site-specific DCE, 
including the projected cost of managing irradiated fuel.
    (f) For decommissioning activities that delay completion of 
decommissioning by including a period of storage or surveillance, the 
licensee must provide a means of adjusting cost estimates and 
associated funding levels over the storage or surveillance period.
    (g) After submitting its site-specific DCE required by paragraph 
(e) of this section, and until the licensee has completed its final 
radiation survey and demonstrated that residual radioactivity has been 
reduced to a level that permits termination of its license, the 
licensee must annually submit to the NRC, by March 31, a financial 
assurance status report. The report must include the following 
information, current through the end of the previous calendar year:
    (1) The amount spent on decommissioning, both cumulative and over 
the previous calendar year, the remaining balance of any 
decommissioning funds, and the amount provided by other financial 
assurance methods being relied upon;
    (2) An estimate of the costs to complete decommissioning, 
reflecting any difference between actual and estimated costs for work 
performed during the year, and the decommissioning criteria upon which 
the estimate is based;
    (3) Any modifications occurring to a licensee's current method of 
providing financial assurance since the last submitted report; and
    (4) Any material changes to trust agreements or financial assurance 
contracts.
    (5) If the sum of the balance of any remaining decommissioning 
funds, plus earnings on such funds calculated at not greater than a 2 
percent real rate of return, together with the amount provided by other 
financial assurance methods being relied upon, does not cover the 
estimated cost to complete the decommissioning, the financial assurance 
status report must include

[[Page 87076]]

additional financial assurance to cover the estimated cost of 
completion.
    (h) After submitting its site-specific DCE required by paragraph 
(e) of this section, the licensee must annually submit to the NRC, by 
March 31, a report on the status of its funding for managing irradiated 
fuel. The report must include the following information, current 
through the end of the previous calendar year:
    (1) The amount of funds accumulated to cover the cost of managing 
the irradiated fuel;
    (2) The projected cost of managing irradiated fuel until title to 
the fuel and possession of the fuel is transferred to the Secretary of 
Energy; and
    (3) If the funds accumulated do not cover the projected cost, a 
plan to obtain additional funds to cover the cost.


Sec.  53.1070  Termination of license.

    For each holder of an OL or COL under this part--
    (a)(1) When the licensee has determined to permanently cease 
operations the licensee must, within 30 days, submit a written 
certification to the NRC, consistent with the requirements of Sec.  
53.040(b)(8);
    (2) When appropriate to support decommissioning activities and the 
eventual permanent removal of fuel from the reactor vessel, the 
licensee must develop defueled technical specifications by reviewing 
the operational technical specifications and determining which 
specifications no longer apply during decommissioning and which ones 
should remain applicable. The licensee must make the appropriate 
submittals to the NRC in accordance with Sec.  53.1510 to request 
changes to the technical specifications; and
    (3)(i) Once fuel has been permanently removed from the reactor 
vessel, the licensee must submit a written certification to the NRC 
that meets the requirements of Sec.  53.040(b)(9); and
    (ii) The licensee must establish and maintain staffing consisting 
of certified fuel handlers, as defined under Sec.  53.020, and other 
non-licensed personnel with appropriate qualifications, and in 
sufficient numbers, to ensure support for facility operations and 
radiological control activities, as required by the facility defueled 
technical specifications. These personnel must be subject to the 
training requirements of Sec.  53.830.
    (b) Upon docketing of the certifications for permanent cessation of 
operations and permanent removal of fuel from the reactor vessel, or 
when a final legally effective order to permanently cease operations 
has come into effect, the license issued under this part no longer 
authorizes operation of the reactor or emplacement or retention of fuel 
into the reactor vessel.
    (c) Decommissioning will be completed within 60 years of permanent 
cessation of operations. Completion of decommissioning beyond 60 years 
will be approved by the Commission only when necessary to protect 
public health and safety. Factors that will be considered by the 
Commission in evaluating an alternative that provides for completion of 
decommissioning beyond 60 years of permanent cessation of operations 
include unavailability of waste disposal capacity and other site-
specific factors affecting the licensee's capability to carry out 
decommissioning, including presence of other nuclear facilities at the 
site.
    (d)(1) Prior to or within 2 years following permanent cessation of 
operations, the licensee must submit a PSDAR and site-specific DCE in 
accordance with Sec.  53.1060(e).
    (2) The NRC must notice receipt of the PSDAR and make the PSDAR 
publicly available and publish notice of its availability for public 
comment in the Federal Register. The NRC must also schedule a public 
meeting readily accessible to individuals in the vicinity of the 
licensee's facility. The NRC must publish a notice in the Federal 
Register and in a forum, such as local newspapers, that is readily 
accessible to individuals in the vicinity of the site, announcing the 
date, time, and location of the meeting, along with a brief description 
of the purpose of the meeting.
    (e) Licensees must not perform any major decommissioning 
activities, as defined in Sec.  53.020, until 90 days after the NRC has 
received the licensee's PSDAR submittal and until certifications of 
permanent cessation of operations and permanent removal of fuel from 
the reactor vessel, as required under paragraph (a) of this section, 
have been submitted.
    (f) Licensees must not perform any decommissioning activities, as 
defined in Sec.  53.020, that--
    (1) Foreclose release of the site for possible unrestricted use;
    (2) Result in significant environmental impacts not previously 
reviewed; or
    (3) Result in there no longer being reasonable assurance that 
adequate funds will be available for decommissioning.
    (g) In taking actions permitted under Sec.  53.1540 following 
submittal of the PSDAR, the licensee must notify the NRC in writing, 
and send a copy to the affected State(s), before performing any 
decommissioning activity inconsistent with, or making any significant 
schedule change from, those actions and schedules described in the 
PSDAR, including changes that increase the decommissioning cost by more 
than 20 percent from the previously provided DCE.
    (h) Licensees may use decommissioning trust funds consistent with 
the limitations of Sec.  53.1045(a). Licensees must report on the 
status of decommissioning trust funds consistent with the requirements 
of Sec.  53.1060.
    (i) Licensees must submit an application for termination of license 
in accordance with Sec.  53.1070. The application for termination of 
license must be accompanied or preceded by a license termination plan 
to be submitted for NRC approval.
    (1) The license termination plan must be a supplement to the Final 
Safety Analysis Report or equivalent and must be submitted at least 2 
years before termination of the license date.
    (2) The license termination plan must include--
    (i) A site characterization;
    (ii) Identification of remaining dismantlement activities;
    (iii) Plans for site remediation;
    (iv) Detailed plans for the final radiation survey;
    (v) A description of the end use of the site, if restricted;
    (vi) An updated site-specific estimate of remaining decommissioning 
costs;
    (vii) A supplement to the environmental report, pursuant to Sec.  
51.53 of this chapter, describing any new information or significant 
environmental change associated with the licensee's proposed 
termination activities; and
    (viii) Identification of parts, if any, of the facility or site 
that were released for use before approval of the license termination 
plan.
    (3) Following receipt of the license termination plan, the NRC must 
make the license termination plan publicly available and publish notice 
of its availability for public comment in the Federal Register. The NRC 
must also schedule a public meeting readily accessible to individuals 
in the vicinity of the licensee's facility upon receipt of the license 
termination plan. The NRC must publish a notice in the Federal Register 
and in a forum, such as local newspapers, that is readily accessible to 
individuals in the vicinity of the site, announcing the date, time, and 
location of the meeting, along with a brief description of the purpose 
of the meeting.
    (j) If the license termination plan demonstrates that the remainder 
of

[[Page 87077]]

decommissioning activities will be performed in accordance with the 
regulations in this chapter, will not be inimical to the common defense 
and security or to the health and safety of the public, and will not 
have a significant effect on the quality of the environment and after 
notice to interested persons, the Commission will approve the plan, by 
license amendment, subject to such conditions and limitations as it 
deems appropriate and necessary and authorize implementation of the 
license termination plan.
    (k) The Commission will terminate the license if it determines 
that--
    (1) The remaining dismantlement has been performed in accordance 
with the approved license termination plan, and
    (2) The final radiation survey and associated documentation, 
including an assessment of dose contributions associated with parts 
released for use before approval of the license termination plan, 
demonstrate that the facility and site have met the criteria for 
decommissioning in subpart E of 10 CFR part 20.


Sec.  53.1075  Program requirements during decommissioning.

    (a) Licensees that have submitted the certifications required under 
Sec.  53.1070 must maintain a decommissioning fire protection program 
to address the potential for fires that could cause the release or 
spread of radioactive materials.
    (1) The objectives of the decommissioning fire protection program 
are to
    (i) Reasonably prevent these fires from occurring;
    (ii) Rapidly detect, control, and extinguish those fires that do 
occur and that could result in a radiological hazard; and
    (iii) Ensure that the risk of fire-induced radiological hazards to 
the public, environment, and plant personnel is minimized.
    (2) The licensee must assess the decommissioning fire protection 
program on a regular basis. The licensee must revise the 
decommissioning fire protection program documentation as appropriate 
throughout the various stages of facility decommissioning.
    (3) The licensee may make changes to the decommissioning fire 
protection program without NRC approval if these changes do not reduce 
the effectiveness of fire protection for structures, systems, and 
components that could result in a radiological hazard, taking into 
account the decommissioning plant conditions and activities.
    (b) [Reserved]


Sec.  53.1080  Release of part of a commercial nuclear plant or site 
for unrestricted use.

    (a) Prior written NRC approval is required to release part of a 
commercial nuclear plant or site for unrestricted use at any time 
before receiving approval of a license termination plan. Section 
53.1060 specifies recordkeeping requirements associated with partial 
release. Holders of an OL or COL under this part seeking NRC review and 
approval must--
    (1) Evaluate the effect of releasing the property to ensure that--
    (i) The dose to individual members of the public does not exceed 
the limits and standards of subpart D of 10 CFR part 20;
    (ii) There is no reduction in the effectiveness of emergency 
planning or physical security;
    (iii) Effluent releases remain within license conditions;
    (iv) The environmental monitoring program and offsite dose 
calculation manual are revised to account for the changes;
    (v) The siting criteria of subpart D of this part continue to be 
met; and
    (vi) All other applicable statutory and regulatory requirements 
continue to be met.
    (2) Perform a historical site assessment of the part of the 
commercial nuclear plant or site to be released; and
    (3) Perform surveys adequate to demonstrate compliance with the 
radiological criteria for unrestricted use specified in Sec.  20.1402 
of this chapter for impacted areas.
    (b) For release of non-impacted areas, the licensee may submit a 
written request for NRC review and approval of the release if a license 
amendment is not otherwise required. The request submittal must 
include--
    (1) The results of the evaluations performed in accordance with 
paragraphs (a)(1) and (a)(2) of this section;
    (2) A description of the part of the commercial nuclear plant or 
site to be released;
    (3) The schedule for release of the property;
    (4) The results of the evaluations performed in accordance with 
Sec.  53.1540; and
    (5) A discussion that provides the reasons for concluding that the 
environmental impacts associated with the licensee's proposed release 
of the property will be bounded by appropriate previously issued 
environmental impact statements.
    (c) After receiving a request from the licensee for NRC approval of 
the release of a non-impacted area, the NRC must--
    (1) Determine whether the licensee has adequately evaluated the 
effect of releasing the property as required by paragraph (a)(1) of 
this section;
    (2) Determine whether the licensee's classification of any release 
areas as non- impacted is adequately justified; and
    (3) If determining that the licensee's submittal is adequate, 
inform the licensee in writing that the release is approved.
    (d) For release of impacted areas, the licensee must submit an 
application for amendment of its license for the release of the 
property. The application must include--
    (1) The information specified in paragraphs (b)(1) through (b)(3) 
of this section;
    (2) The methods used for and results obtained from the radiation 
surveys required to demonstrate compliance with the radiological 
criteria for unrestricted use specified in Sec.  20.1402; and
    (3) A supplement to the environmental report, under Sec.  51.53 of 
this chapter, describing any new information or significant 
environmental change associated with the licensee's proposed release of 
the property.
    (e) After receiving a license amendment application from the 
licensee for the release of an impacted area, the NRC must--
    (1) Determine whether the licensee has adequately evaluated the 
effect of releasing the property as required by paragraph (a)(1) of 
this section;
    (2) Determine whether the licensee's classification of any release 
areas as non-impacted is adequately justified;
    (3) Determine whether the licensee's radiation survey for an 
impacted area is adequate; and
    (4) If determining that the licensee's submittal is adequate, 
approve the licensee's amendment application.
    (f) The NRC must publish notice receipt of the release approval 
request or license amendment application in the Federal Register and 
make the approval request or license amendment application available 
for public comment. Before acting on an approval request or license 
amendment application submitted in accordance with this section, the 
NRC must conduct a public meeting readily accessible to individuals in 
the vicinity of the licensee's facility for the purpose of obtaining 
public comments on the proposed release of part of the commercial 
nuclear plant or site. The NRC must publish a document in the Federal 
Register and in a forum, such as local newspapers, which is readily 
accessible to individuals in the vicinity of the site, announcing the 
date, time,

[[Page 87078]]

and location of the meeting, along with a brief description of the 
purpose of the meeting.

Subpart H--Licenses, Certifications, and Approvals


Sec.  53.1100  Filing of application for licenses, certifications, or 
approvals; oath or affirmation.

    (a) Serving of applications. (1) Each filing of an application for 
a standard design approval, standard design certification, or license 
under this part, and any amendments to the applications, must be 
submitted to the U.S. Nuclear Regulatory Commission (NRC) under Sec.  
53.040, as applicable.
    (2) Each applicant for a construction permit (CP), early site 
permit, combined license (COL), or manufacturing license (ML) under 
this part must, upon notification by the presiding officer designated 
to conduct the public hearing required by the Atomic Energy Act of 
1954, as amended, (the Act) update the application and serve the 
updated copies of the application or parts of it, eliminating all 
superseded information, together with an index of the updated 
application, as directed by presiding officer. Any subsequent amendment 
to the application must be served on those served copies of the 
application and must be submitted to the NRC as specified in Sec.  
53.040, as applicable.
    (3) The applicant must make a copy of the updated application 
available at the public hearing for the use of any other parties to the 
proceeding and must certify that the updated copies of the application 
contain the current contents of the application submitted in accordance 
with the requirements under this part.
    (4) At the time of filing an application, the Commission will make 
available at the NRC website, http://www.nrc.gov, a copy of the 
application, subsequent amendments, and other records pertinent to the 
matter that is the subject of the application for public inspection and 
copying.
    (5) The serving of copies required by this section must not occur 
until the application has been docketed under Sec.  2.101(a) of this 
chapter. Copies must be submitted to the Commission, as specified in 
Sec.  53.040, as applicable, to enable the Director, Office of Nuclear 
Reactor Regulation to determine whether the application is sufficiently 
complete to permit docketing.
    (b) Oath or affirmation. Each application for a standard design 
approval, standard design certification, or license, including, 
whenever appropriate, a CP or early site permit, or amendment of it, 
and each amendment of each application must be executed in a signed 
original by the applicant or duly authorized officer thereof under oath 
or affirmation.
    (c) [Reserved]
    (d) [Reserved]
    (e) Filing fees. Each application for a standard design approval, 
standard design certification, or commercial nuclear plant license 
under this part, including, whenever appropriate, a CP, COL, operating 
license (OL), ML, or early site permit, other than a license exempted 
from 10 CFR part 170, must be accompanied by the fee prescribed in 10 
CFR part 170. No fee will be required to accompany an application for 
renewal, amendment, or termination of a CP, OL, COL, or ML, except as 
provided in Sec.  170.21 of this chapter.
    (f) Environmental report. An application for a CP, OL, early site 
permit, design certification, COL, or ML for a commercial nuclear plant 
must be accompanied by an environmental report required under subpart A 
of 10 CFR part 51.


Sec.  53.1101  Requirement for license.

    Except as provided in Sec.  53.1120, no person within the United 
States may transfer or receive in interstate commerce, manufacture, 
produce, transfer, acquire, possess, or use any utilization facility 
except as authorized by a license issued by the Commission.


Sec.  53.1103  Combining applications and licenses.

    (a) An applicant may combine several applications in one 
application for different kinds of licenses under the regulations in 
this chapter.
    (b) The Commission may combine in a single license the activities 
of an applicant which would otherwise be licensed separately.


Sec.  53.1106  Elimination of repetition.

    An applicant may incorporate by reference in its application 
information contained in previous applications, statements, or reports 
filed with the Commission, provided, however, that such references are 
clear and specific.


Sec.  53.1109  Contents of applications; general information.

    Each application must include, unless otherwise indicated in this 
subpart--
    (a) Name of applicant;
    (b) Address of applicant;
    (c) Description of business or occupation of applicant;
    (d)(1) If applicant is an individual, the citizenship of applicant;
    (2) If applicant is a partnership, the name, citizenship and 
address of each partner and the principal location where the 
partnership does business;
    (3) If applicant is a corporation or an unincorporated association, 
the following information:
    (i) The State where it is incorporated or organized and the 
principal location where it does business;
    (ii) The names, addresses and citizenship of its directors and of 
its principal officers; and
    (iii) Whether it is owned, controlled, or dominated by an alien, a 
foreign corporation, or foreign government, and if so, give details; or
    (4) If the applicant is acting as agent or representative of 
another person in filing the application, identify the principal and 
furnish information required under this paragraph with respect to such 
principal;
    (e) The class and type of license applied for, the use to which the 
facility will be put, the period of time for which the license is 
sought, and a list of other licenses, except operator's licenses, 
issued or applied for in connection with the proposed facility;
    (f) [Reserved]
    (g)(1) Except as provided in paragraph (g)(2) of this section, if 
the application is for an OL or COL for a commercial nuclear plant, or 
if the application is for an early site permit for a commercial nuclear 
plant and contains plans for coping with emergencies under Sec.  
53.1146(b)(2)(ii), the applicant must submit the radiological emergency 
response plans of State, local, and participating Tribal governmental 
entities in the United States that are wholly or partially within the 
plume exposure pathway emergency planning zone (EPZ),\1\ and the plans 
of State governments wholly or partially within the ingestion pathway 
EPZ.\2\ If the application is for an early site permit that, under 
Sec.  53.1146(b)(2)(i), proposes major features of the emergency plans 
describing the EPZs, then the descriptions of the EPZs must meet the 
requirements of this paragraph. Generally, the plume exposure pathway 
EPZ for a commercial nuclear plant must consist of an area about 10 
miles (16 km) in radius and the ingestion pathway EPZ must consist of 
an area about 50 miles (80 km) in radius. The exact size and 
configuration of the EPZs surrounding a particular commercial nuclear 
plant must be determined in relation to the local emergency response 
needs and capabilities as they are affected by such conditions as 
demography, topography, land characteristics, access routes, and 
jurisdictional boundaries. The size of the EPZs also may be determined 
on a case-by-case basis for gas-cooled reactors and for reactors with 
an authorized power level less than 250

[[Page 87079]]

megawatt thermal. The plans for the ingestion pathway must focus on 
such actions as are appropriate to protect the food ingestion pathway.
    (2) Applicants for commercial nuclear plants consisting of either 
small modular reactors or non-light-water reactors complying with Sec.  
50.160 of this chapter who apply for a CP, an OL, a COL, or an early 
site permit under this part must submit as part of the application the 
analysis used to determine whether the criteria in Sec.  
53.1109(g)(2)(i)(A) and (B) are met and, if they are met, the size of 
the plume exposure pathway EPZ.
    (i) The plume exposure pathway EPZ is the area within which:
    (A) Public dose, as defined in Sec.  20.1003 of this chapter, is 
projected to exceed 10 millisieverts (1 rem) total effective dose 
equivalent over 96 hours from the release of radioactive materials from 
the facility considering accident likelihood and source term, timing of 
the accident sequence, and meteorology; and
    (B) Pre-determined, prompt protective measures are necessary.
    (ii) If the application is for an OL or COL or if the application 
is for an early site permit and contains plans for coping with 
emergencies under Sec.  53.1146(b)(2)(ii), and if the plume exposure 
pathway EPZ extends beyond the site boundary:
    (A) The applicant must submit radiological emergency response plans 
of State, local, and participating Tribal governmental entities in the 
United States that are wholly or partially within the plume exposure 
pathway EPZ.
    (B) The exact configuration of the plume exposure pathway EPZ 
surrounding the facility shall be determined in relation to the local 
emergency response needs and capabilities as they are affected by such 
conditions as demography, topography, land characteristics, access 
routes, and jurisdictional boundaries.
    (iii) If the application is for an early site permit that, under 
Sec.  53.1146(b)(2)(i), proposes major features of the emergency plans 
and describes the EPZ, and if the EPZ extends beyond the site boundary, 
then the exact configuration of the plume exposure pathway EPZ 
surrounding the facility must be determined in relation to the local 
emergency response needs and capabilities as they are affected by such 
conditions as demography, topography, land characteristics, access 
routes, and jurisdictional boundaries.
    (h) [Reserved]
    (i) A list of the names and addresses of such regulatory agencies 
as may have jurisdiction over the rates and services incident to the 
proposed activity, and a list of trade and news publications which 
circulate in the area where the proposed activity will be conducted and 
which are considered appropriate to give reasonable notice of the 
application to those municipalities, private utilities, public bodies, 
and cooperatives, which might have a potential interest in the 
facility; and
    (j) If the application contains Restricted Data or classified 
National Security information, confirmation that all Restricted Data 
and classified National Security information are separated from the 
unclassified information.

    \1\ EPZs are discussed in NUREG-0396, U.S. Environmental 
Protection Agency 520/1-78-016, ``Planning Basis for the Development 
of State and Local Government Radiological Emergency Response Plans 
in Support of Light-Water Nuclear Power Plants,'' December 1978.
    \2\ If the State, local, and participating Tribal emergency 
response plans have been previously provided to the NRC for 
inclusion in the facility docket, the applicant need only provide 
the appropriate reference to meet this requirement.


Sec.  53.1112  Environmental conditions.

    (a) Each CP, early site permit, and COL under this part may include 
conditions to address environmental issues during construction. These 
conditions are to be set out in an attachment to the license, which is 
incorporated in and made a part of the license. These conditions will 
be derived from information contained in the environmental report 
submitted pursuant to Sec.  51.50 of this chapter, as analyzed and 
evaluated in the NRC record of decision and will identify the 
obligations of the licensee in the environmental area, including, as 
appropriate, requirements for reporting and keeping records of 
environmental data, and any conditions and monitoring requirement for 
the protection of the nonaquatic environment.
    (b) Each license authorizing operation of a commercial nuclear 
plant under this part, and each license for a commercial nuclear plant 
for which the certification of permanent cessation of operations 
required under Sec.  53.1070 has been submitted may include conditions 
to address environmental issues during operation and decommissioning. 
These conditions are to be set out in an attachment to the license, 
which is incorporated in and made a part of the license. These 
conditions will be derived from information contained in the 
environmental report or the supplement to the environmental report 
submitted under Sec. Sec.  51.50 and 51.53 of this chapter as analyzed 
and evaluated in the NRC record of decision, and will identify the 
obligations of the licensee in the environmental area, including, as 
appropriate, requirements for reporting and keeping records of 
environmental data and any conditions and monitoring requirement for 
the protection of the nonaquatic environment.


Sec.  53.1115  Agreement limiting access to classified information.

    As part of its application and in any event before the receipt of 
Restricted Data or classified National Security Information or the 
issuance of a license or standard design approval under this part, or 
before the Commission has adopted a final standard design certification 
rule under this part, the applicant must agree in writing that it will 
not permit any individual to have access to or any facility to possess 
Restricted Data or classified National Security Information until the 
individual and/or facility has been approved for access under the 
provisions of 10 CFR parts 25 and/or 95. The agreement of the applicant 
becomes part of the license or standard design approval.


Sec.  53.1118  Ineligibility of certain applicants.

    Any person who is a citizen, national, or agent of a foreign 
country, or any corporation, or other entity which the Commission knows 
or has reason to believe is owned, controlled, or dominated by an 
alien, a foreign corporation, or a foreign government, will be 
ineligible to apply for and obtain a license.


Sec.  53.1120  Exceptions and exemptions from licensing requirements.

    Nothing in this part must be deemed to require a license for--
    (a) The manufacture, production, or acquisition by the Department 
of Defense of any utilization facility authorized pursuant to section 
91 of the Act or the use of such facility by the Department of Defense 
or by a person under contract with and for the account of the 
Department of Defense;
    (b) Except to the extent that the Department of Energy facilities 
of the types subject to licensing pursuant to section 202 of the Energy 
Reorganization Act of 1974, as amended, are involved--
    (1)(i) The processing, fabrication or refining of special nuclear 
material (SNM) or the separation of SNM, or the separation of SNM from 
other substances by a prime contractor of the Department of Energy 
under a prime contract for--

[[Page 87080]]

    (A) The performance of work for the Department of Energy at a 
United States government-owned or controlled site;
    (B) Research in, or development, manufacture, storage, testing or 
transportation of, atomic weapons or components thereof; or
    (C) The use or operation of a utilization facility in a United 
States owned vehicle or vessel; or
    (ii) The processing, fabrication or refining of SNM of the 
separation of SNM, or the separation of SNM from other substances by a 
prime contractor or subcontractor of the Commission or the Department 
of Energy under a prime contract or subcontract when the Commission 
determines that the exemption of the prime contractor or subcontractor 
is authorized by law; and that, under the terms of the contract or 
subcontract, there is adequate assurance that the work thereunder can 
be accomplished without undue risk to the public health and safety; or
    (2)(i) The construction or operation of a utilization facility for 
the Department of Energy at a United States government-owned or 
controlled site, including the transportation of the utilization 
facility to or from such site and the performance of contract services 
during temporary interruptions of such transportation; or the 
construction or operation of a utilization facility for the Department 
of Energy in the performance of research in, or development, 
manufacture, storage, testing, or transportation of, atomic weapons or 
components thereof; or the use or operation of a utilization facility 
for the Department of Energy in a United States government-owned 
vehicle or vessel; provided that such activities are conducted by a 
prime contractor of the Department of Energy under a prime contract 
with the Department of Energy; or
    (ii) The construction or operation of a utilization facility by a 
prime contractor or subcontractor of the Commission or the Department 
of Energy under his or her prime contract or subcontract when the 
Commission determines that the exemption of the prime contractor or 
subcontractor is authorized by law; and that, under the terms of the 
contract or subcontract, there is adequate assurance that the work 
thereunder can be accomplished without undue risk to the public health 
and safety; or
    (c) The transportation or possession of any utilization facility by 
a common or contract carrier or warehouse employee in the regular 
course of carriage for another or storage incident thereto.


Sec.  53.1121  Public inspection of applications.

    Applications and documents submitted to the Commission in 
connection with applications may be made available for public 
inspection under the provisions of part 2 of this chapter.


Sec.  53.1124  Relationship between sections.

    (a) Limited work authorization. An application for a limited work 
authorization (LWA) under this part may be submitted as part of an 
application for an early site permit, CP, or COL under this part as 
required in Sec.  53.1130(a)(2).
    (b) Early site permit. (1) A holder of an early site permit may 
request an LWA.
    (2) An application for a CP or COL under this part may, but need 
not, reference an early site permit.
    (c) Standard design approval. An application for a standard design 
approval under this part may, but need not, reference an OL or custom 
COL under this part that is essentially the same as the information 
supporting the standard design for which approval is being requested.
    (d) Standard design certification. An application for a standard 
design certification under this part may, but need not, reference an OL 
or custom COL under this part that is essentially the same as the 
standard design for which certification is being requested.
    (e) Manufacturing license. (1) A manufactured reactor manufactured 
under an ML issued under this part may only be transported to and 
installed at a site for which a COL under this part has been issued.
    (2) An ML applicant under this part may reference a standard design 
certification or a standard design approval under this part in its 
application.
    (f) Construction permit. An application for a CP may, but need not, 
reference a standard design certification or standard design approval 
issued under this part, respectively, and may also reference an early 
site permit issued under this part. In the absence of a demonstration 
that an entity other than the one originally sponsoring a standard 
design certification is qualified to supply a design, the Commission 
will entertain an application for a CP that references a standard 
design certification issued under this part only if the entity that 
sponsored the certification supplies the design for the applicant's 
use.
    (g) Operating license. (1) An application for an OL under this part 
may, but need not, reference an early site permit, standard design 
certification, or standard design approval issued under this part. In 
the absence of a demonstration that an entity other than the one 
originally sponsoring a standard design certification is qualified to 
supply a design, the Commission will entertain an application for an OL 
that references a standard design certification issued under this part 
only if the entity that sponsored the certification supplies the design 
for the applicant's use.
    (2) The holder of a CP must, at the time of submission of the Final 
Safety Analysis Report (FSAR), file an application for an OL.
    (h) Combined licenses. An application for a COL under this part 
may, but need not, reference an early site permit, standard design 
certification, standard design approval, or ML issued under this part. 
In the absence of a demonstration that an entity other than the one 
originally sponsoring and obtaining a standard design certification is 
qualified to supply a design, the Commission will entertain an 
application for a COL that references a standard design certification 
issued under this part only if the entity that sponsored the 
certification supplies the design for the applicant's use.


Sec.  53.1130  Limited work authorizations.

    (a) Request for limited work authorization. (1) Any person to whom 
the Commission may otherwise issue either a license or permit related 
to a commercial nuclear plant may request an LWA allowing that person 
to perform the driving of piles, subsurface preparation, placement of 
backfill, concrete, or permanent retaining walls within an excavation, 
and installation of the foundation, including placement of concrete, 
any of which are for a structure, system, or component (SSC) of the 
facility for which either a CP or COL is otherwise required under Sec.  
53.610.
    (2) An application for an LWA may be submitted as part of a 
complete application for a CP or COL in accordance with Sec.  
2.101(a)(1) through (a)(5) of this chapter, or as a partial application 
in accordance with Sec.  2.101(a)(9) of this chapter. An application 
for an LWA by the holder of an early site permit must be submitted as a 
complete application in accordance with Sec.  2.101(a)(1) through 
(a)(4) of this chapter.
    (3) The application must include--
    (i) A Safety Analysis Report required by Sec. Sec.  53.1146, 
53.1309 or 53.1416, as applicable, a description of the activities 
requested to be performed, and the design and construction information

[[Page 87081]]

otherwise required by the Commission's rules and regulations to be 
submitted for a CP or COL under this part but limited to those portions 
of the facility that are within the scope of the LWA. The Safety 
Analysis Report must demonstrate that activities conducted under the 
LWA will be conducted in compliance with the technically relevant 
Commission requirements in 10 CFR chapter I applicable to the design of 
those portions of the facility within the scope of the LWA;
    (ii) An environmental report in accordance with Sec.  51.49 of this 
chapter; and
    (iii) A plan for redress of activities performed under the LWA, 
should limited work activities be terminated by the holder, or the LWA 
be revoked by the NRC or upon effectiveness of the Commission's final 
decision denying the associated CP or COL application, as applicable.
    (b) Issuance of limited work authorization. (1) The Director, 
Office of Nuclear Reactor Regulation may issue an LWA only after--
    (i) The NRC staff issues the final environmental impact statement 
for the LWA under subpart A of part 51 of this chapter;
    (ii) The presiding officer makes the finding in Sec. Sec.  
51.105(c) or 51.107(d) of this chapter, as applicable;
    (iii) The Director determines that the applicable standards and 
requirements of the Act, and the Commission's regulations applicable to 
the activities to be conducted under the LWA, have been met, the 
applicant is technically qualified to engage in the activities 
authorized, and that issuance of the LWA will provide reasonable 
assurance of adequate protection to public health and safety and will 
not be inimical to the common defense and security; and
    (iv) The presiding officer finds that there are no unresolved 
safety issues relating to the activities to be conducted under the LWA 
that would constitute good cause for withholding the authorization.
    (2) Each LWA will specify the activities that the holder is 
authorized to perform.
    (c) Effect of limited work authorization. Any activities undertaken 
under an LWA are entirely at the risk of the applicant and, except as 
to the matters determined under paragraph (b)(1) of this section, the 
issuance of the LWA has no bearing on the issuance of a CP or COL with 
respect to the requirements of the Act and rules, regulations, or 
orders issued under the Act. The environmental impact statement for a 
CP or COL application for which an LWA was previously issued will not 
address, and the presiding officer will not consider, the sunk costs of 
the holder of the LWA in determining the proposed action (i.e., 
issuance of the CP or COL).
    (d) Implementation of redress plan. If construction is terminated 
by the holder, the underlying application is withdrawn by the applicant 
or denied by the NRC, or the LWA is revoked by the NRC, then the holder 
must begin implementation of the redress plan in a reasonable time. The 
holder must also complete the redress of the site no later than 18 
months after termination of construction, revocation of the LWA, or 
upon effectiveness of the Commission's final decision denying the 
associated CP application or the associated COL application, as 
applicable.


Sec.  53.1140  Early site permits.

    Sections 53.1140 through 53.1188 set out the requirements and 
procedures applicable to Commission issuance of an early site permit 
under this part for approval of a site for a commercial nuclear plant 
separate from the filing of an application for a CP or COL for the 
facility.


Sec.  53.1143  Filing of applications.

    Any person who may apply for a CP or for a COL under this part, may 
file an application for an early site permit with the Director, Office 
of Nuclear Reactor Regulation. An application for an early site permit 
may be filed notwithstanding the fact that an application for a CP or a 
COL has not been filed in connection with the site for which a permit 
is sought.


Sec.  53.1144  Contents of applications for early site permits; general 
information.

    The application must contain all of the information required by 
Sec.  53.1109(a) through (d) and (j).


Sec.  53.1146  Contents of applications for early site permits; 
technical information.

    (a) The application must contain--
    (1) A Site Safety Analysis Report that must include the following:
    (i) The specific number, type, and thermal power level of the 
facilities, or range of possible facilities, for which the site may be 
used;
    (ii) The anticipated maximum levels of radiological and thermal 
effluents each facility will produce;
    (iii) The type of cooling systems, including intakes and outflows, 
where appropriate, that may be associated with each facility;
    (iv) The boundaries of the site;
    (v) The proposed general location of each facility on the site;
    (vi) The external hazards and site characteristics required by this 
part;
    (vii) The location and description of any nearby industrial, 
military, or transportation facilities and routes;
    (viii) The existing and projected future population profile of the 
area surrounding the site;
    (ix) A description and assessment of the site on which a facility 
is to be located. The assessment must address the requirements of 
subpart D of this part;
    (x) Information demonstrating that site characteristics are such 
that adequate security plans and measures can be developed; and
    (xi) A description of the quality assurance program (QAP) required 
by appendix B to part 50 of this chapter applied to site-related 
activities for the future design, fabrication, construction, and 
testing of the SSCs of a facility or facilities that may be constructed 
on the site.
    (2) A complete environmental report as required by Sec.  51.50(b) 
of this chapter.
    (b)(1) The Site Safety Analysis Report must identify physical 
characteristics of the proposed site, such as egress limitations from 
the area surrounding the site, that could pose a significant impediment 
to the development of emergency plans. If physical characteristics are 
identified that could pose a significant impediment to the development 
of emergency plans, the application must identify measures that would, 
when implemented, mitigate or eliminate the significant impediment.
    (2) The Site Safety Analysis Report may also--
    (i) Propose major features of the emergency plans, under either 
Sec.  50.160 or the requirements in appendix E to part 50 and Sec.  
50.47(b) of this chapter, as applicable, such as the exact size and 
configuration of the EPZs, for review and approval by the NRC, in 
consultation with the Federal Emergency Management Agency (FEMA), as 
applicable, in the absence of complete and integrated emergency plans; 
or
    (ii) Propose complete and integrated emergency plans for review and 
approval by the NRC, in consultation with FEMA, as applicable, in 
accordance with either Sec.  50.160 or the requirements in appendix E 
to part 50 and Sec.  50.47(b) of this chapter. To the extent approval 
of emergency plans is sought, the application must contain the 
information required by Sec.  53.1109(g).
    (3) Emergency plans submitted under paragraph (b)(2)(ii) of this 
section must include the proposed inspections, tests, and analyses that 
the holder of a COL referencing the early site permit must perform, and 
the acceptance criteria that are necessary and sufficient to provide

[[Page 87082]]

reasonable assurance that, if the inspections, tests, and analyses are 
performed and the acceptance criteria met, the facility has been 
constructed and will be operated in conformity with the emergency 
plans, the provisions of the Act, and the Commission's rules and 
regulations. Major features of an emergency plan submitted under 
paragraph (b)(2)(i) of this section may include proposed inspections, 
tests, analyses, and acceptance criteria (ITAAC).
    (4) Under paragraphs (b)(1) and (b)(2)(i) of this section, the Site 
Safety Analysis Report must include, where appropriate, a description 
of contacts and arrangements made with Federal, State, participating 
Tribal, and local governmental agencies with emergency planning 
responsibilities. The Site Safety Analysis Report must contain any 
certifications that have been obtained. If these certifications, where 
appropriate, cannot be obtained, the Site Safety Analysis Report must 
contain information, including a utility plan, sufficient to show that 
the proposed plans provide reasonable assurance that adequate 
protective measures can and will be taken in the event of a 
radiological emergency at the site. Under the option set forth in 
paragraph (b)(2)(ii) of this section, the applicant must make good 
faith efforts, where appropriate, to obtain from the same governmental 
agencies certifications that--
    (i) The proposed emergency plans are practicable;
    (ii) These agencies are committed to participating in any further 
development of the plans, including any required field demonstrations; 
and
    (iii) That these agencies are committed to executing their 
responsibilities under the plans in the event of an emergency.
    (c) An applicant may request that an LWA under Sec.  53.1130 be 
issued in conjunction with the early site permit. The application must 
include the information otherwise required by Sec.  53.1130.
    (d) Each applicant for an early site permit under this part must 
protect safeguards information against unauthorized disclosure in 
accordance with the requirements in Sec. Sec.  73.21 and 73.22 of this 
chapter, as applicable.


Sec.  53.1149  Review of applications.

    (a) Standards for review of applications. Applications filed under 
this part will be reviewed according to the applicable standards set 
out in this part. In addition, the Commission must prepare an 
environmental impact statement during review of the application, under 
the applicable provisions of 10 CFR part 51. The Commission must 
determine, after consultation with FEMA, as applicable, whether the 
information required of the applicant by Sec.  53.1146(b)(1) shows that 
there is no significant impediment to the development of emergency 
plans that cannot be mitigated or eliminated by measures proposed by 
the applicant, whether any major features of emergency plans submitted 
by the applicant under Sec.  53.1146(b)(2)(i) are acceptable under 
either Sec.  50.160 or appendix E to part 50 and Sec.  50.47(b) of this 
chapter, and whether any emergency plans submitted by the applicant 
under Sec.  53.1146(b)(2)(ii) provide reasonable assurance that 
adequate protective measures can and will be taken in the event of a 
radiological emergency.
    (b) Administrative review of applications; hearings. An early site 
permit application is subject to all procedural requirements in 10 CFR 
part 2, including the requirements for docketing in Sec.  2.101(a)(1) 
through (4) of this chapter, and the requirements for issuance of a 
notice of hearing in Sec.  2.104(a) and (d) of this chapter, provided 
that the designated sections may not be construed to require that the 
environmental report, or draft or final environmental impact statement 
includes an assessment of the benefits of construction and operation of 
the reactor or reactors, or an analysis of alternative energy sources. 
The presiding officer in an early site permit hearing must not admit 
contentions proffered by any party concerning an assessment of the 
benefits of construction and operation of the reactor or reactors, or 
an analysis of alternative energy sources if those issues were not 
addressed by the applicant in the early site permit application. All 
hearings conducted on applications for early site permits filed under 
this part are governed by the procedures contained in subparts C, G, L, 
and N of 10 CFR part 2, as applicable.


Sec.  53.1155  Referral to the Advisory Committee on Reactor 
Safeguards.

    The Commission must refer a copy of the application for an early 
site permit to the Advisory Committee on Reactor Safeguards (ACRS). The 
ACRS must report on those portions of the application which concern 
safety.


Sec.  53.1158  Issuance of early site permit.

    (a) After conducting a hearing under Sec.  53.1149(b) and receiving 
the report to be submitted by the ACRS under Sec.  53.1155, the 
Commission may issue an early site permit, in the form the Commission 
deems appropriate, if the Commission finds that--
    (1) An application for an early site permit demonstrates compliance 
with the applicable standards and requirements of the Act and the 
Commission's regulations;
    (2) Notifications, if any, to other agencies or bodies have been 
duly made;
    (3) There is reasonable assurance that the site is in conformity 
with the provisions of the Act and the Commission's regulations;
    (4) The applicant is technically qualified to engage in any 
activities authorized;
    (5) The proposed ITAAC, including any on emergency planning, are 
necessary and sufficient, within the scope of the early site permit, to 
provide reasonable assurance that the facility has been constructed and 
will be operated in conformity with the license, the provisions of the 
Act, and the Commission's regulations;
    (6) Issuance of the permit will not be inimical to the common 
defense and security or to the health and safety of the public;
    (7) Any significant adverse environmental impact resulting from 
activities requested under Sec.  53.1146(c) can be redressed; and
    (8) The findings required by subpart A of 10 CFR part 51 have been 
made.
    (b) The early site permit must specify the site characteristics, 
design parameters, and terms and conditions of the early site permit 
the Commission deems appropriate. Before issuance of either a CP or COL 
referencing an early site permit, the Commission must find that any 
relevant terms and conditions of the early site permit have been met. 
Any terms or conditions of the early site permit that could not be met 
by the time of issuance of the CP or COL, must be set forth as terms or 
conditions of the CP or COL.
    (c) The early site permit must specify those Sec.  53.1130(b) 
activities requested under Sec.  53.1146(c) that the permit holder is 
authorized to perform.


Sec.  53.1161  Extent of activities permitted.

    If the activities authorized by Sec.  53.1158(c) are performed and 
the site is not referenced in an application for a CP or a COL issued 
under this part while the permit remains valid, then the early site 
permit remains in effect solely for the purpose of site redress, and 
the holder of the permit must redress the site under the terms of the 
site redress plan required by Sec.  53.1146(c). If, before redress is 
complete, a use not envisaged in the redress plan is found for the site 
or parts thereof, the holder of the permit

[[Page 87083]]

must carry out the redress plan to the greatest extent possible 
consistent with the alternate use.


Sec.  53.1164  Duration of permit.

    (a) Except as provided in paragraph (b) of this section, an early 
site permit issued under this subpart may be valid for not less than 
10, nor more than 20 years from the date of issuance.
    (b) An early site permit continues to be valid beyond the date of 
expiration in any proceeding on a CP application or a COL application 
that references the early site permit and is docketed before the date 
of expiration of the early site permit, or, if a timely application for 
renewal of the permit has been docketed, before the Commission has 
determined whether to renew the permit.
    (c) An applicant for a CP or COL may, at its own risk, reference in 
its application a site for which an early site permit application has 
been docketed but not granted.
    (d) Upon issuance of a CP or COL, a referenced early site permit is 
subsumed, to the extent referenced, into the CP or COL.


Sec.  53.1167  Limited work authorization after issuance of early site 
permit.

    A holder of an early site permit may request an LWA under Sec.  
53.1130.


Sec.  53.1170  Transfer of early site permit.

    An application to transfer an early site permit will be processed 
under Sec.  53.1570.


Sec.  53.1173  Application for renewal.

    (a) Not less than 12, nor more than 36 months before the expiration 
date stated in the early site permit, or any later renewal period, the 
permit holder may apply for a renewal of the permit. An application for 
renewal must contain all information necessary to bring up to date the 
information and data contained in the previous application.
    (b) Any person whose interests may be affected by renewal of the 
permit may request a hearing on the application for renewal. The 
request for a hearing must comply with Sec.  2.309 of this chapter. If 
a hearing is granted, notice of the hearing will be published under 
Sec.  2.309 of this chapter.
    (c) An early site permit, either original or renewed, for which a 
timely application for renewal has been filed, remains in effect until 
the Commission has determined whether to renew the permit. If the 
permit is not renewed, it continues to be valid in certain proceedings 
in accordance with the provisions of Sec.  53.1164(b).
    (d) The Commission must refer a copy of the application for renewal 
to the ACRS. The ACRS must report on those portions of the application 
which concern safety and must apply the criteria set forth in Sec.  
53.1176.


Sec.  53.1176  Criteria for renewal.

    (a) The Commission must grant the renewal if it determines that--
    (1) The site complies with the Act, the Commission's regulations, 
and orders applicable and in effect at the time the site permit was 
originally issued; and
    (2) Any new requirements the Commission may wish to impose--
    (i) Are necessary for adequate protection to public health and 
safety or common defense and security;
    (ii) Are necessary for compliance with the Commission's 
regulations, and orders applicable and in effect at the time the site 
permit was originally issued; or
    (iii) Would provide a substantial increase in overall protection of 
the public health and safety or the common defense and security to be 
derived from the new requirements, and the direct and indirect costs of 
implementation of those requirements are justified in view of this 
increased protection.
    (b) A denial of renewal under the provisions of Sec.  53.1176(a) 
does not bar the permit holder or another applicant from filing a new 
application for the site which proposes changes to the site or the way 
that it is used to correct the deficiencies cited in the denial of the 
renewal.


Sec.  53.1179  Duration of renewal.

    Each renewal of an early site permit may be for not less than 10, 
nor more than 20 years, plus any remaining years on the early site 
permit then in effect before renewal.


Sec.  53.1182  Use of site for other purposes.

    A site for which an early site permit has been issued under this 
part may be used for purposes other than those described in the permit, 
including the location of other types of energy facilities. The permit 
holder must inform the Director, Office of Nuclear Reactor Regulation 
(Director), of any significant uses for the site which have not been 
approved in the early site permit. The information about the activities 
must be given to the Director at least 30 days in advance of any actual 
construction or site modification for the activities. The information 
provided could be the basis for imposing new requirements on the 
permit, under the provisions of Sec.  53.1188. If the permit holder 
informs the Director that the holder no longer intends to use the site 
for a commercial nuclear plant, the Director may terminate the permit.


Sec.  53.1188  Finality of early site permit determinations.

    (a) Commission finality. (1) While an early site permit is in 
effect under Sec.  53.1164 or Sec.  53.1179, the Commission may not 
change or impose new site characteristics, design parameters, or terms 
and conditions, including emergency planning requirements, on the early 
site permit unless the Commission--
    (i) Determines that a modification is necessary to bring the permit 
or the site into compliance with the Commission's regulations and 
orders applicable and in effect at the time the permit was issued;
    (ii) Determines the modification is necessary to assure adequate 
protection of the public health and safety or the common defense and 
security;
    (iii) Determines that a modification is necessary based on an 
update under paragraph (b) of this section; or
    (iv) Issues a variance requested under paragraph (d) of this 
section.
    (2) In making the findings required for issuance of a CP, COL, or 
OL, or the findings required by Sec.  53.1452(g), or in any enforcement 
hearing other than one initiated by the Commission under paragraph 
(a)(1) of this section, if the application for the CP, COL, or OL 
references an early site permit, the Commission must treat as resolved 
those matters resolved in the proceeding on the application for 
issuance or renewal of the early site permit, except as provided for in 
paragraphs (b), (c), and (d) of this section.
    (i) If the Commission grants a CP application that references an 
early site permit and an application for an OL references the CP, the 
Commission must treat as resolved those matters resolved in the 
proceeding for the issuance or renewal of the early site permit, except 
as provided for in paragraphs (b), (c), and (d) of this section.
    (ii) If the early site permit approved an emergency plan (or major 
features thereof) that is in use by a licensee of a commercial nuclear 
plant, the Commission must treat as resolved changes to the early site 
permit emergency plan (or major features thereof) that are identical to 
changes made to the licensee's emergency plans under Sec.  53.1565 
occurring after issuance of the early site permit.
    (iii) If the early site permit approved an emergency plan (or major 
features thereof) that is not in use by a licensee of a commercial 
nuclear plant, the Commission must treat as resolved changes that are 
equivalent to those that could be made under Sec.  53.1565 without 
prior NRC approval had the emergency plan been in use by a licensee.

[[Page 87084]]

    (b) Updating of early site permit-emergency preparedness. An 
applicant for a CP, OL, or COL who has filed an application referencing 
an early site permit issued under this subpart must update the 
emergency preparedness information that was provided under Sec.  
53.1146(b) and discuss whether the updated information materially 
changes the bases for compliance with applicable NRC requirements.
    (c) Hearings and petitions. (1) In any proceeding for the issuance 
of a CP, OL, or COL referencing an early site permit, contentions on 
the following matters may be litigated in the same manner as other 
issues material to the proceeding:
    (i) The nuclear reactor proposed to be built does not fit within 
one or more of the site characteristics or design parameters included 
in the early site permit;
    (ii) One or more of the terms and conditions of the early site 
permit have not been met;
    (iii) A variance requested under paragraph (d) of this section is 
unwarranted or should be modified;
    (iv) New or additional information is provided in the application 
that substantially alters the bases for a previous NRC conclusion or 
constitutes a sufficient basis for the Commission to modify or impose 
new terms and conditions related to emergency preparedness; or
    (v) Any significant environmental issue that was not resolved in 
the early site permit proceeding, or any issue involving the impacts of 
construction and operation of the facility that was resolved in the 
early site permit proceeding for which significant new information has 
been identified.
    (2) Any person may file a petition requesting that the site 
characteristics, design parameters, or terms and conditions of the 
early site permit be modified, or that the permit be suspended or 
revoked. The petition will be considered under Sec.  2.206 of this 
chapter. Before construction commences, the Commission must consider 
the petition and determine whether any immediate action is required. If 
the petition is granted, an appropriate order will be issued. 
Construction under the CP or COL will not be affected by the granting 
of the petition unless the order is made immediately effective. Any 
change required by the Commission in response to the petition must 
demonstrate compliance with the requirements of paragraph (a)(1) of 
this section.
    (d) Variances. An applicant for a CP, OL, or COL referencing an 
early site permit may include in its application a request for a 
variance from one or more site characteristics, design parameters, or 
terms and conditions of the early site permit, or from the Site Safety 
Analysis Report. In determining whether to grant the variance, the 
Commission must apply the same technically relevant criteria applicable 
to the application for the original or renewed early site permit. Once 
a CP or COL referencing an early site permit is issued, variances from 
the early site permit will not be granted for that CP or COL.
    (e) Early site permit amendment. The holder of an early site permit 
may not make changes to the early site permit or the Site Safety 
Analysis Report without prior Commission approval. The request for a 
change to the early site permit must be in the form of an application 
for a license amendment and must demonstrate compliance with the 
requirements of Sec. Sec.  53.1510 and 53.1520.


Sec.  53.1200  Standard design approvals.

    Sections 53.1200 through 53.1221 set out procedures for the filing, 
NRC staff review, and referral to the ACRS of standard designs, or 
major portions thereof, for a commercial nuclear plant under this part.


Sec.  53.1203  Filing of applications.

    Any person may submit a proposed standard design for a commercial 
nuclear plant to the NRC staff for its review. The submittal may 
consist of either the final design for the entire facility or the final 
design for major portions thereof.


Sec.  53.1206  Contents of applications for standard design approvals; 
general information.

    The application must contain all of the information required by 
Sec.  53.1109(a) through (c) and (j).


Sec.  53.1209  Contents of applications for standard design approvals; 
technical information.

    (a) Major portion of a standard design. If the applicant seeks 
review of a major portion of a standard design, the application need 
only contain the information required by this section to the extent the 
requirements are applicable to the major portion of the standard design 
for which NRC staff approval is sought. If an applicant seeks approval 
of a major portion of the design, the scope of the application for 
which approval is sought must include all functional design criteria 
necessary to demonstrate compliance with the safety criteria in 
Sec. Sec.  53.210, 53.220 and 53.450(e), as applicable, for the major 
portion of the standard design for which NRC staff approval is sought. 
Such applicants must identify conditions related to interfaces with 
systems outside the scope of the major portion of the standard design 
for which NRC staff approval is sought, and functional or physical 
boundary conditions between the major portion of the standard design 
for which NRC staff approval is sought and the remainder of the 
standard design. These conditions must be demonstrated when the 
standard design approval is incorporated into a subsequent CP, design 
certification, ML, or COL application.
    (b) Final Safety Analysis Report. The application must contain an 
FSAR that describes the facility and the limits on its operation, 
presents a safety analysis of the SSCs and of the facility, or major 
portions thereof, for which the applicant seeks design approval, and 
must include the following information:
    (1) Site Parameters. The site parameters postulated for the design 
under this part, including the design-basis external hazard levels for 
the relevant external hazards, and an analysis and evaluation of the 
design in terms of those site parameters.
    (2) Design information. Except as specified in this paragraph, an 
application for a standard design approval for a commercial nuclear 
plant must include the design information equivalent to that required 
for a standard design certification under Sec.  53.1239(a)(2) through 
(27) for those portions of a commercial nuclear plant included in the 
standard design approval.


Sec.  53.1210  Contents of applications for standard design approvals; 
other application content.

    (a) In addition to the FSAR, the application must also include the 
following:
    (1) Availability Controls (if not included in the FSAR). A 
description of the controls on plant operations, including availability 
controls, to provide reasonable confidence that the configurations and 
special treatments for safety-related (SR) SSCs and non-safety-related 
but safety-significant (NSRSS) SSCs provide the capabilities and 
reliabilities required to demonstrate compliance with the safety 
criteria of Sec.  53.220.
    (2) Safeguards Information. A description of the program to protect 
Safeguards Information against unauthorized disclosure in accordance 
with the requirements in Sec. Sec.  73.21 and 73.22 of this chapter, as 
applicable.
    (b) If there are SSCs of the plant which required research and 
development to confirm the adequacy of their design, provide a report 
in the application which documents the

[[Page 87085]]

resolution of any safety questions associated with such SSCs.
    (c) A description of how the performance of each design feature has 
been demonstrated capable of fulfilling functional design criteria 
considering interdependent effects through either analysis, appropriate 
test programs, prototype testing, operating experience, or a 
combination thereof, in accordance with Sec.  53.440(a).


Sec.  53.1212  Standards for review of applications.

    Applications filed under this part will be reviewed under the 
standards set out in 10 CFR parts 20, 53, and 73.


Sec.  53.1215  Referral to the Advisory Committee on Reactor 
Safeguards.

    The Commission must refer a copy of the application to the ACRS. 
The ACRS must report on those portions of the application which concern 
safety.


Sec.  53.1218  Staff approval of design.

    (a) Upon completion of its review of a submittal under Sec. Sec.  
53.1200 through 53.1221 and receipt of a report by the ACRS under Sec.  
53.1215, the NRC staff must publish a determination in the Federal 
Register as to whether or not the design is acceptable, subject to 
appropriate terms and conditions, and make an analysis of the design in 
the form of a report available at the NRC website, https://www.nrc.gov.
    (b) A standard design approval issued under this section is valid 
for 15 years from the date of issuance and may not be renewed. A design 
approval continues to be valid beyond the date of expiration in any 
proceeding on an application for a CP, OL, COL, or ML under this part 
that references the design approval and is docketed before the date of 
expiration of the design approval.


Sec.  53.1221  Finality of standard design approvals; information 
requests.

    (a) An approved design must be used by and relied upon by the NRC 
staff and the ACRS in their reviews of any standard design 
certification or individual facility license application under this 
part that incorporates by reference a standard design approved under 
this part unless there exists significant new information that 
substantially affects the earlier determination or other good cause.
    (b) The determination and report by the NRC staff do not constitute 
a commitment to issue a permit or license, or in any way affect the 
authority of the Commission, Atomic Safety and Licensing Board Panel, 
or presiding officers in any proceeding under part 2 of this chapter.
    (c) Except for information requests seeking to verify compliance 
with the current licensing basis of the standard design approval, 
information requests to the holder of a standard design approval must 
be evaluated before issuance to ensure that the burden to be imposed on 
respondents is justified in view of the potential safety significance 
of the issue to be addressed in the requested information. Each 
evaluation performed by the NRC staff must be in accordance with Sec.  
53.1580 and must be approved by the Executive Director for Operations 
or authorized designee before issuance of the request.
    (d) The Commission will require, before granting a CP, COL, OL, or 
ML that references a standard design approval, that engineering 
documents, such as analyses, drawings, procurement specifications, or 
construction and installation specifications, be completed and 
available for audit if the more detailed information is necessary for 
the Commission to verify the information in the application and make 
its safety determination, including the determination that the 
application is consistent with the design approval information. This 
information may be acquired by appropriate arrangements with the design 
approval applicant.


Sec.  53.1230  Standard design certifications.

    Sections 53.1230 through 53.1263 set forth the requirements and 
procedures applicable to the Commission's issuance of rules granting 
standard design certifications for commercial nuclear plants under this 
part separate from the filing of an application for a CP or COL for 
such a facility.


Sec.  53.1233  Filing of applications.

    (a) An application for design certification may be filed 
notwithstanding the fact that an application for a CP, COL, or ML for 
such a facility has not been filed.
    (b) The application must comply with the applicable filing 
requirements of Sec.  53.040 and Sec. Sec.  2.811 through 2.819 of this 
chapter.


Sec.  53.1236  Contents of applications for standard design 
certifications; general information.

    The application must contain all of the information required by 
Sec.  53.1109(a) through (c) and (j).


Sec.  53.1239  Contents of applications for standard design 
certifications; technical information.

    The application must contain a level of design information 
sufficient to enable the Commission to judge the applicant's proposed 
means of assuring that construction conforms to the design and to reach 
a final conclusion on all safety questions associated with the design 
before the certification is granted. The information submitted for a 
design certification must include performance requirements and design 
information sufficiently detailed to permit the preparation of 
acceptance and inspection requirements by the NRC. The Commission will 
require, before design certification, that information normally 
contained in engineering documents, such as analyses, drawings, 
procurement specifications, or construction and installation 
specifications, be completed and available for audit if the more 
detailed information is necessary for the Commission to verify the 
information in the application and make its safety determination.
    (a) Final Safety Analysis Report. The application must contain an 
FSAR that describes the facility and the limits on its operation, and 
presents a safety analysis of the SSCs, and must include the following 
information:
    (1) Site Parameters. The site parameters postulated for the design 
under this part, including the design-basis external hazard levels for 
the relevant external hazards, and an analysis and evaluation of the 
design in terms of those site parameters.
    (2) Plant Description and Safety Functions--(i) General Plant 
Description. A general description of the commercial nuclear plant 
including reactor type, the intended use of the reactor, nuclear design 
(e.g., neutron spectrum, reactor control, multi-unit reactor control), 
overall layout of the plant including significant plant features and 
SSCs, maximum power level and the nature and inventory of radioactive 
materials.
    (ii) Safety functions. A description of the primary and additional 
safety functions required under Sec.  53.230 and a summary of how each 
safety function is satisfied.
    (3) Design Features and functional design criteria--licensing-basis 
events. (i) A description of the design features required by Sec.  
53.400 and the functional design criteria required by Sec. Sec.  53.410 
and 53.420 that, when combined with corresponding human actions and 
programmatic controls, demonstrate that the plant will demonstrate 
compliance with the safety criteria defined in Sec.  53.210 and 
established in accordance with Sec.  53.220, or more restrictive 
alternative criteria adopted under Sec.  53.470, during licensing-basis 
events (LBEs).
    (ii) A description of how design features demonstrate compliance 
with

[[Page 87086]]

the requirements of Sec.  53.440(a) through (i) and (k) through (m).
    (4) Design Features Supporting Normal Operations. A description of 
the design features required by Sec.  53.425 to support the holder of 
an OL or COL complying with Sec.  53.260 during normal operations.
    (5) Design Features and Functional Design Criteria--aircraft 
impact. A description of the design features and functional design 
criteria required to demonstrate compliance with the requirements of 
Sec.  53.440(j) for addressing the impact of a large, commercial 
aircraft.
    (6) Earthquake engineering. The information necessary to 
demonstrate that the commercial nuclear plant complies with the 
earthquake engineering criteria in Sec.  53.480.
    (7) Programmatic Controls and Interfaces. (i) A description of the 
corresponding programmatic controls and interfaces necessary to achieve 
and maintain the reliability and capability of SSCs relied upon to 
demonstrate compliance with the functional design criteria required by 
Sec. Sec.  53.410 and 53.420 and the safety criteria in Sec. Sec.  
53.210 and 53.220, or more restrictive alternative criteria adopted 
under Sec.  53.470, and necessary to maintain consistency with analyses 
required by Sec.  53.450.
    (ii) For an application for a multi-unit commercial nuclear plant, 
the programmatic controls and interfaces must also be described for 
different modular configurations, as required by Sec.  53.440(i), 
including any restrictions that will be necessary during the 
construction and startup of any given unit to ensure the safe operation 
of the overall commercial nuclear plant to be licensed under this part.
    (8) Programmatic Controls for Normal Operations. A description of 
how programmatic controls, including monitoring programs, would provide 
assurance that design features and procedures will enable the holder of 
an OL or COL to comply with Sec.  53.260.
    (9) Design Features Supporting the Protection of Plant Workers. A 
description of the design features required by Sec.  53.430 to support 
the holder of an OL or COL complying with Sec.  53.270.
    (10) Programmatic Controls for Protection of Plant Workers. A 
description of how programmatic controls, including monitoring 
programs, would provide assurance that design features and procedures 
will enable the holder of an OL or COL to comply with Sec.  53.270.
    (11) Codes and Standards. A description of generally accepted 
consensus codes and standards used to design the design features, as 
required by Sec.  53.440(b).
    (12) Materials. A description of the materials used for SR and 
NSRSS SSCs and a description of the qualification of these materials 
for their service conditions over the plant lifetime, as required by 
Sec.  53.440(c).
    (13) Integrity Assessment Program. A description of a design 
integrity assessment program that addresses the elements described in 
Sec.  53.440(d).
    (14) Safety and Security. Confirmation that safety and security 
were considered together in the design process, as required by Sec.  
53.440(f).
    (15) Criticality. Information demonstrating how the applicant will 
comply with requirements for criticality accidents in Sec.  53.440(m).
    (16) Multi-unit Plants. For an application for standard design 
certification of a multi-unit commercial nuclear plant, the possible 
operating configurations of the reactor units, including common 
systems, interface requirements, and system interactions, as required 
by Sec.  53.440(i).
    (17) SSC Classification. (i) The classification of SSCs according 
to their safety significance under Sec.  53.460(a).
    (ii) For SR and NSRSS SSCs, the conditions under which they must 
perform the safety functions required by Sec.  53.230, including 
environmental conditions.
    (18) Probabilistic Risk Assessment. A description of the 
probabilistic risk assessment (PRA) required by Sec.  53.450(a) and its 
results.
    (19) Analyses. A description of the analyses performed under Sec.  
53.450(b) through (g) that includes the following information:
    (i) A description of the analysis of LBEs and its results, as 
described in Sec.  53.240. This analysis description must--
    (A) Address the elements in Sec.  53.450(e) and (f); and
    (B) Under Sec.  53.460(c)--
    (1) Describe any human actions that are necessary to prevent or 
mitigate LBEs;
    (2) Describe how those human actions are capable of being reliably 
performed under the postulated environmental conditions present; and
    (3) Describe how those human actions would be addressed by programs 
established under subpart F of this part.
    (ii)(A) A description of how SSCs relied on to meet the safety 
criteria defined in Sec.  53.210 are protected against or designed to 
withstand the effects of external hazards under Sec.  53.510.
    (B) The information necessary to demonstrate that the commercial 
nuclear plant complies with the earthquake engineering criteria in 
Sec.  53.480.
    (iii) A description of the defense-in-depth measures required by 
Sec.  53.250.
    (iv) A description of all plant operating states where there is the 
potential for the uncontrolled release of radioactive material to the 
environment, as required by Sec.  53.450(b)(4).
    (v) A description of the events that challenge plant control and 
safety systems whose failure could lead to an undesirable end state 
and/or radioactive material release, as required by Sec.  53.450(b)(5).
    (vi) A description of the analytical codes used in modeling plant 
behavior in analyses of LBEs and how these codes are qualified for the 
range of conditions for which they were used, as required by Sec.  
53.450(d).
    (vii) If not described in addressing paragraph (5) of this section, 
the results of other analyses required by Sec.  53.450(g).
    (20) Special Treatments. A description of special treatments 
established as required by Sec.  53.460.
    (21) Analytical Margins. A description of any alternative criteria 
adopted to demonstrate analytical margins supporting operational 
flexibilities, if applicable, as required by Sec.  53.470.
    (22) Quality Assurance. A description of the QAP applied to the 
design of the SSCs of the commercial nuclear plant, as required by 
Sec.  53.460(b). The description of the QAP for a commercial nuclear 
plant must include a discussion of how the applicable requirements of 
appendix B to part 50 of this chapter were satisfied.
    (23) Design Features and Controls to Address the Minimization of 
Contamination. The information required by Sec.  20.1406 of this 
chapter.
    (24) Interface Requirements. (i) A description analysis, and 
evaluation of the interfaces between the standard design and the 
balance of the commercial nuclear plant that may impact the ability of 
the plant to demonstrate compliance with the functional design criteria 
or the safety criteria of subparts B and C of this part.
    (ii) Confirmation that interface requirements are verifiable 
through inspections, testing, or analysis. These requirements must be 
sufficiently detailed to allow for completion of the final safety 
analysis by license applicants that reference the certified design 
under this subpart. The method to be used for verification of interface 
requirements must be included as part of the proposed ITAAC required by 
Sec.  53.1241(a)(3).

[[Page 87087]]

    (iii) A representative conceptual design for those portions of the 
plant for which the application does not seek certification to aid the 
NRC in its review of the FSAR and to permit assessment of the adequacy 
of the interface requirements under paragraph (a)(24)(i) of this 
section.
    (25) Technical Qualifications. A description of the technical 
qualifications of the applicant to engage in the proposed activities in 
accordance with the regulations in this chapter.
    (26) Technical Specifications. Proposed technical specifications 
prepared under Sec.  53.710(a) for those areas addressed by the design 
certification.
    (27) Role of personnel. Information to address the following for 
the role of personnel in ensuring safe operations:
    (i) A description of how the human factors engineering design 
requirements of Sec.  53.440(n)(1) are addressed;
    (ii) A description of how the human system interface design 
requirements of Sec.  53.440(n)(2) are addressed;
    (iii) A concept of operations that is of sufficient scope and 
detail to address the requirements of Sec.  53.440(n)(3);
    (iv) A functional requirements analysis and function allocation 
that is of sufficient scope and detail to address the requirements of 
Sec.  53.440(n)(4).
    (b) [Reserved]


Sec.  53.1241  Contents of applications for standard design 
certifications; other application content.

    (a) In addition to the FSAR, the application must also include the 
following:
    (1) Environmental report. An environmental report as required by 
Sec.  51.55 of this chapter.
    (2) Availability Controls (if not included in the FSAR). A 
description of the controls on plant operations, including availability 
controls, to provide reasonable confidence that the configurations and 
special treatments for SR and NSRSS SSCs provide the capabilities and 
reliabilities required to demonstrate compliance with the safety 
criteria of Sec.  53.220, or more restrictive alternative criteria 
adopted under Sec.  53.470.
    (3) Inspections, tests, analyses, and acceptance criteria. The 
proposed ITAAC that are necessary and sufficient to provide reasonable 
assurance that, if the inspections, tests, and analyses are performed 
and the acceptance criteria met, a facility that incorporates the 
design certification has been constructed and will be operated in 
conformity with the design certification, the provisions of the Act, 
and the Commission's rules and regulations.
    (4) Safeguards information. A description of the program to protect 
Safeguards Information against unauthorized disclosure in accordance 
with the requirements in Sec. Sec.  73.21 and 73.22 of this chapter, as 
applicable.
    (b) If there are SSCs of the plant which required research and 
development to confirm the adequacy of their design, provide a report 
in the application which documents the resolution of any safety 
questions associated with such SSCs.
    (c) A description of how the performance of each design feature has 
been demonstrated capable of fulfilling functional design criteria 
considering interdependent effects through either analysis, appropriate 
test programs, prototype testing, operating experience, or a 
combination thereof, in accordance with Sec.  53.440(a).


Sec.  53.1242  Review of applications.

    (a) Standards for review of applications. Applications filed under 
this part will be reviewed for compliance with the standards set out in 
this part and 10 CFR parts 20, 51, and 73.
    (b) Administrative review of applications; hearings. (1) A standard 
design certification is a rule that will be issued under the provisions 
of subpart H of 10 CFR part 2, as supplemented by the provisions of 
this section. The Commission must initiate the rulemaking after an 
application has been filed under Sec.  53.1233 and must specify the 
procedures to be used for the rulemaking. The notice of proposed 
rulemaking published in the Federal Register must provide an 
opportunity for the submission of comments on the proposed design 
certification rule. If, at the time a proposed design certification 
rule is published in the Federal Register under this paragraph, the 
Commission decides that a legislative hearing should be held, the 
information required by Sec.  2.1502(c) of this chapter must be 
included in the Federal Register document for the proposed design 
certification.
    (2) Following the submission of comments on the proposed design 
certification rule, the Commission may, at its discretion, hold a 
legislative hearing under the procedures in subpart O of part 2 of this 
chapter. The Commission must publish a document in the Federal Register 
of its decision to hold a legislative hearing. The document must 
contain the information specified in Sec.  2.1502(c) of this chapter 
and specify whether the Commission or a presiding officer will conduct 
the legislative hearing.
    (3) Notwithstanding anything in Sec.  2.390 of this chapter to the 
contrary, proprietary information will be protected in the same manner 
and to the same extent as proprietary information submitted in 
connection with applications for licenses, provided that the design 
certification will be published in chapter I of this title.
    (c) Reference to an issued operating license or combined license. 
In those cases where a design certification application is preceded by 
the issuance of an OL or custom COL for a commercial nuclear plant that 
is essentially the same as the standard design for which certification 
is being requested, the NRC review will follow the processes for 
referencing a standard design approval in Sec.  53.1221, to the extent 
practicable.


Sec.  53.1245  Referral to the Advisory Committee on Reactor 
Safeguards.

    The Commission must refer a copy of the application to the ACRS. 
The ACRS must report on those portions of the application which concern 
safety.


Sec.  53.1248  Issuance of standard design certification.

    (a) After conducting a rulemaking proceeding under Sec.  53.1242 on 
an application for a standard design certification and receiving the 
report to be submitted by the ACRS under Sec.  53.1245, the Commission 
may issue a standard design certification in the form of a rule for the 
design that is the subject of the application, if the Commission 
determines that--
    (1) The application demonstrates compliance with the applicable 
standards and requirements of the Act and the Commission's regulations;
    (2) Notifications, if any, to other agencies or bodies have been 
duly made;
    (3) There is reasonable assurance that the standard design conforms 
with the provisions of the Act and the Commission's regulations;
    (4) The applicant is technically qualified;
    (5) The proposed ITAAC are necessary and sufficient, within the 
scope of the standard design, to provide reasonable assurance that, if 
the inspections, tests, and analyses are performed and the acceptance 
criteria met, the facility has been constructed and will be operated in 
accordance with the design certification, the provisions of the Act, 
and the Commission's regulations;
    (6) Issuance of the standard design certification will not be 
inimical to the common defense and security or to the health and safety 
of the public;

[[Page 87088]]

    (7) The findings required by subpart A of part 51 of this chapter 
have been made; and
    (8) The applicant has implemented the QAP described or referenced 
in the Safety Analysis Report.
    (b) The design certification rule must specify the site parameters, 
design characteristics, and any additional requirements and 
restrictions of the design certification rule.
    (c) After the Commission has adopted a final design certification 
rule, the applicant must not permit any individual to have access to or 
any facility to possess restricted data or classified National Security 
Information until the individual and/or facility has been approved for 
access under the provisions of 10 CFR parts 25 and/or 95, as 
applicable.


Sec.  53.1251  Duration of certification.

    (a) Except as provided in paragraph (b) of this section, a standard 
design certification issued under this subpart is valid for 15 years 
from the effective date of the rule.
    (b) A standard design certification continues to be valid beyond 
the date of expiration in any proceeding on an application for a COL or 
an OL under this part that references the standard design certification 
and is docketed either before the date of expiration of the 
certification, or, if a timely application for renewal of the 
certification has been filed, before the Commission has determined 
whether to renew the certification. A design certification also 
continues to be valid beyond the date of expiration in any hearing held 
under Sec.  53.1452 before operation begins under a COL that references 
the design certification.
    (c) An applicant for a CP, OL, COL, or ML under this part may, at 
its own risk, reference in its application a design for which a design 
certification application has been docketed but not granted.


Sec.  53.1254  Application for renewal.

    (a) Not less than 12 nor more than 36 months before the expiration 
of the initial 15-year period, or any later renewal period, any person 
may apply for renewal of the certification. An application for renewal 
must contain all information necessary to bring up to date the 
information and data contained in the previous application. The 
Commission will require, before renewal of certification, that 
engineering documents, such as analyses, drawings, procurement 
specifications, or construction and installation specifications, be 
completed and available for audit if the more detailed information is 
necessary for the Commission to verify the information in the 
application and make its safety determination. Notice and comment 
procedures must be used for a rulemaking proceeding on the application 
for renewal. The Commission, in its discretion, may require the use of 
additional procedures in individual renewal proceedings.
    (b) A design certification, either original or renewed, for which a 
timely application for renewal has been filed remains in effect until 
the Commission has determined whether to renew the certification. If 
the certification is not renewed, it continues to be valid in certain 
proceedings under Sec.  53.1251.
    (c) The Commission must refer a copy of the application for renewal 
to the ACRS. The ACRS must report on those portions of the application 
which concern safety and must apply the criteria set forth in Sec.  
53.1257.


Sec.  53.1257  Criteria for renewal.

    (a) The Commission must issue a rule granting the renewal if the 
design, either as originally certified or as modified during the 
rulemaking on the renewal, complies with the Act and the Commission's 
regulations applicable and in effect at the time the certification was 
issued.
    (b) The Commission may impose other requirements if it determines 
that--
    (1) They are necessary for adequate protection to public health and 
safety or common defense and security;
    (2) They are necessary for compliance with the Commission's 
regulations and orders applicable and in effect at the time the design 
certification was issued; or
    (3) There is a substantial increase in overall protection of the 
public health and safety or the common defense and security to be 
derived from the new requirements, and the direct and indirect costs of 
implementing those requirements are justified in view of this increased 
protection.
    (c) In addition, the applicant for renewal may request an amendment 
to the design certification. The Commission must grant the amendment 
request if it determines that the amendment will comply with the Act 
and the Commission's regulations in effect at the time of renewal. If 
the amendment request entails such an extensive change to the design 
certification that an essentially new standard design is being 
proposed, an application for a design certification must be filed in 
accordance with this subpart.
    (d) Denial of renewal does not bar the applicant, or another 
applicant, from filing a new application for certification of the 
design, which proposes design changes that correct the deficiencies 
cited in the denial of the renewal.


Sec.  53.1260  Duration of renewal.

    Each renewal of certification for a standard design will be for not 
less than 10, nor more than 15 years.


Sec.  53.1263  Finality of standard design certifications.

    (a)(1) While a standard design certification rule is in effect 
under Sec. Sec.  53.1251 or 53.1260, the Commission may not modify, 
rescind, or impose new requirements on the certification information, 
whether on its own motion, or in response to a petition from any 
person, unless the Commission determines in a rulemaking that the 
change--
    (i) Is necessary either to bring the certification information or 
the referencing plants into compliance with the Commission's 
regulations applicable and in effect at the time the certification was 
issued;
    (ii) Is necessary to provide adequate protection of the public 
health and safety or the common defense and security;
    (iii) Reduces unnecessary regulatory burden and maintains 
protection to public health and safety and the common defense and 
security;
    (iv) Provides the detailed design information to be verified under 
those ITAAC that are directed at certification information (i.e., 
design acceptance criteria);
    (v) Is necessary to correct material errors in the certification 
information;
    (vi) Substantially increases overall safety, reliability, or 
security of facility design, construction, or operation, and the direct 
and indirect costs of implementation of the rule change are justified 
in view of this increased safety, reliability, or security; or
    (vii) Contributes to increased standardization of the certification 
information.
    (2)(i) In a rulemaking under Sec.  53.1263(a)(1), except for Sec.  
53.1263(a)(1)(ii), the Commission will give consideration to whether 
the benefits justify the costs for plants that are already licensed or 
for which an application for a permit or license is under 
consideration.
    (ii) The rulemaking procedures for changes under Sec.  
53.1263(a)(1) must provide for notice and opportunity for public 
comment.
    (3) Any modification the NRC imposes on a design certification rule

[[Page 87089]]

under paragraph (a)(1) of this section will be applied to all plants 
referencing the certified design, except those to which the 
modification has been rendered technically irrelevant by action taken 
under paragraphs (a)(4) or (b) of this section.
    (4) The Commission may not impose new requirements by plant-
specific order on any part of the design of a specific plant 
referencing the design certification rule if that part was approved in 
the design certification while a design certification rule is in effect 
under Sec.  53.1248, unless--
    (i) A modification is necessary to secure compliance with the 
Commission's regulations applicable and in effect at the time the 
certification was issued, or to assure adequate protection of the 
public health and safety or the common defense and security; and
    (ii) Special circumstances as defined in Sec.  53.080 are present. 
In addition to the factors listed in Sec.  53.080, the Commission must 
consider whether the special circumstances which Sec.  53.080 requires 
to be present outweigh any decrease in safety that may result from the 
reduction in standardization caused by the plant-specific order.
    (5) Except as provided in Sec.  2.335 of this chapter, in making 
the findings required for issuance of a COL, CP, OL, or ML, or for any 
hearing under Sec.  53.1452, the Commission must treat as resolved 
those matters resolved in connection with the issuance or renewal of a 
design certification rule.
    (b) An applicant who references a design certification rule may 
request an exemption from one or more elements of the certification 
information. The Commission may grant such a request only if it 
determines that the exemption will comply with the requirements of 
Sec.  53.080. In addition to the factors listed in Sec.  53.080, the 
Commission must consider whether the special circumstances that Sec.  
53.080 requires to be present outweigh any decrease in safety that may 
result from the reduction in standardization caused by the exemption. 
The granting of an exemption on request of an applicant is subject to 
litigation in the same manner as other issues in the OL or COL hearing.
    (c) The Commission will require, before granting a CP, COL, OL, or 
ML that references a design certification rule, that information 
normally contained in engineering documents, such as analyses, 
drawings, procurement specifications, or construction and installation 
specifications, be completed and available for audit if the more 
detailed information is necessary for the Commission to verify the 
information in the application and make its safety determination, 
including the determination that the application is consistent with the 
certification information. This information may be acquired by 
appropriate arrangements with the design certification applicant.


Sec.  53.1270  Manufacturing licenses.

    Sections 53.1270 through 53.1295 set out the requirements and 
procedures applicable to Commission issuance of a license under this 
part authorizing manufacture of manufactured reactors to be installed 
at sites not identified in the ML application.


Sec.  53.1273  Filing of applications.

    Any person, except one excluded by Sec.  53.1118, may file an 
application for an ML under this part with the Director, Office of 
Nuclear Reactor Regulation.


Sec.  53.1276  Contents of applications for manufacturing licenses; 
general information.

    Each application for an ML must include the information contained 
in Sec.  53.1109(a) through (e), and (j).


Sec.  53.1279  Contents of applications for manufacturing licenses; 
technical information.

    (a) Final Safety Analysis Report-siting and design. The application 
must include an FSAR containing the information set forth below, with a 
level of design information sufficient to enable the Commission to 
judge the applicant's proposed means of ensuring that the manufacturing 
conforms to the design and to reach a final conclusion on all safety 
questions associated with the design, permit the preparation of 
construction and installation specifications by an applicant who seeks 
to use the manufactured reactor, and permit the preparation of 
acceptance and inspection requirements by the NRC. The application must 
include the following information:
    (1) Site parameters. The site parameters postulated for the design 
under this part, including the design-basis external hazard levels for 
the relevant external hazards, and an analysis and evaluation of the 
design in terms of those site parameters.
    (2) Design information. Except as specified in this paragraph, the 
design information equivalent to that required for a standard design 
certification as defined in Sec.  53.1239(a)(2) through (27) for those 
portions of a commercial nuclear plant included in the manufactured 
reactor.
    (3) Quality assurance program. A description of the QAP applied to 
the design, and to be applied to the fabrication and testing of the 
SSCs of the manufactured reactor under Sec.  53.620(a)(6), including a 
discussion of how the applicable requirements of appendix B to part 50 
of this chapter will be satisfied;
    (4) Conceptual designs. Representative conceptual designs for one 
or more commercial nuclear plants using the manufactured reactor;
    (5) Operating configurations. If multiple manufactured reactors may 
be installed at a commercial nuclear plant, a description of the 
possible operating configurations, including common systems, interface 
requirements, and system interactions. The final safety analysis must 
also account for differences among the possible configurations, 
including any restrictions that will be necessary during the 
construction and startup of a given manufactured reactor to ensure the 
safe operation of any commercial nuclear reactor already operating;
    (6) Interface requirements. (i) The interface requirements between 
the manufactured reactor and the remaining portions of the commercial 
nuclear plant or connections to other facilities outside of the 
commercial nuclear plant.
    (ii) Confirmation that interface requirements are verifiable 
through inspections, testing, or analysis. These requirements must be 
sufficiently detailed to allow for completion of the final safety 
analysis by license applicants that reference the manufactured reactor 
manufactured under this subpart. Applicants for a COL under this part 
will need to verify the interface requirements at the installation 
site. The method to be used for verification of interface requirements 
must be included as part of the proposed ITAAC required by Sec.  
53.1282(a).
    (iii) Information to support development of radiation monitoring 
programs required under subpart F of this part by an applicant for a 
COL, including potential pathways for radionuclides produced within the 
manufactured reactor to enter interfacing systems.
    (b) Final Safety Analysis Report--manufacturing information. The 
FSAR must include the following information related to the 
manufacturing processes, organization, controls, and inspections:
    (1) A description, including references to generally accepted 
consensus codes and standards, of the processes that will be used to 
procure, fabricate, and assemble components that make up the 
manufactured reactor. The description should clearly define which 
activities are proposed to be within the

[[Page 87090]]

scope of the ML and those, such as the making of a component to be 
procured from a separate company for installation in the manufactured 
reactor, that are not considered to be within the scope of the ML;
    (2) A description of the organizational and management structure 
singularly responsible for direction of design and manufacture of the 
manufactured reactor. The information should include a description of 
the management plans, technical qualifications, and controls in place 
to demonstrate compliance with the requirements of Sec.  53.620;
    (3) A description of the inspections and tests to be performed as 
part of the manufacturing process, including the inspection of procured 
components, inspection and testing of fabrication processes such as the 
molding, welding, or coating of components, and inspections and testing 
of the assembled manufactured reactor or portions of the manufactured 
reactor;
    (4) A description of the fitness-for-duty program required by part 
26 of this chapter and its implementation.
    (c) Deployment of the completed manufactured reactor. The 
application must include the following information related to the 
deployment of a manufactured reactor:
    (1) Procedures governing the preparation of the manufactured 
reactor or portions of the manufactured reactor for shipping to the 
site where it is to be operated; the conduct of shipping; and verifying 
the condition of the shipped items upon receipt at the site;
    (2) Details of the interaction of the design, manufacture, and 
installation of a manufactured reactor within the applicant's 
organization and the manner by which the applicant will ensure close 
integration between the designer, contractors, and any facility in 
which the manufactured reactor is to be installed;
    (3) A description of the measures to be used for the control of 
interfaces, including the consideration of key site parameters, between 
the holder of the ML and the holder of the COL for the commercial 
nuclear plant at which the manufactured reactor is to be installed.
    (d) Special considerations for factory fueling. In addition to the 
above paragraphs, an application for an ML for a manufactured reactor 
that will be fueled at the factory under a 10 CFR part 70 license must 
include the following information related to loading fuel and the 
required independent physical mechanisms to prevent criticality and to 
otherwise provide assurance that the fueled manufactured reactor can be 
successfully transported, installed, and operated at a site for which 
the Commission has issued a COL that authorizes construction and 
operation of a commercial nuclear plant using the manufactured reactor:
    (1) A description of the procedures used during the fueling of the 
manufactured reactor that ensure that the configuration of fuel within 
the fueled manufactured reactor is consistent with the design and 
analyses supporting operation of the manufactured reactor under the COL 
at the place of operation. The description may reference the applicable 
10 CFR part 70 application and other sections of the Safety Analysis 
Report supporting the ML license application.
    (i) The application must describe the measures taken for in-factory 
inspections and non-nuclear testing performed to ensure that the 
configuration of fuel within the fueled manufactured reactor is 
consistent with the design and analyses supporting operation of the 
manufactured reactor under the COL at the place of operation.
    (ii) The application must describe the design features included in 
the manufactured reactor to prevent criticality, including at least two 
independent mechanisms each of which is sufficient to prevent 
criticality, the associated functional design criteria applied to those 
design features, and the physical and programmatic controls implemented 
during manufacturing, storage, and transport that are credited to 
assure the features function as designed when subject to potential 
hazards and human errors. The descriptions must include how those 
measures will be controlled during installation under the ML and 
removal under the COL at the place of operation.
    (2) A description of the procedures governing the transfer of 
responsibilities for the fueled manufactured reactor from the holder of 
the ML to the holder of the COL for the installation site.
    (3) If available at the time of filing the ML application or, if 
not available at the time of filing the ML application, submitted as an 
amendment to the ML or ML application at the time of filing the Part 70 
application, a description of the programs needed to demonstrate 
compliance with the requirements of Sec.  53.620(d) and 10 CFR parts 
70, 71, and 73 for the receipt, storage, and loading of SNM into a 
manufactured reactor and the transport of the fueled manufactured 
reactor to a site for which the Commission has issued a COL that 
authorizes construction and operation of a commercial nuclear plant 
using the manufactured reactor, including the following.
    (i) A physical security program in accordance with Sec.  
53.620(d)(2)(i).
    (ii) A cybersecurity program in accordance with Sec.  
53.620(d)(2)(i).


Sec.  53.1282  Contents of applications for manufacturing licenses; 
other application content.

    (a) Inspections, tests, analyses, and acceptance criteria. (1) The 
application must contain proposed inspections, tests, and analyses that 
the COL holder must perform, and the acceptance criteria that are 
necessary and sufficient to provide reasonable assurance that, if the 
inspections, tests, and analyses are performed and the acceptance 
criteria met:
    (i) The reactor has been manufactured in conformity with the ML, 
the provisions of the Act, and the Commission's rules and regulations; 
and
    (ii) The manufactured reactor will be operated in conformity with 
the approved design and any license authorizing operation of the 
manufactured reactor.
    (2) If the application references a standard design certification, 
the ITAAC contained in the certified design must apply to those 
portions of the facility design that are covered by the design 
certification.
    (3) If the application references a standard design certification, 
the application may include a notification that a required inspection, 
test, or analysis in the design certification ITAAC has been 
successfully completed and that the corresponding acceptance criterion 
has been met. The Federal Register notification required by Sec.  
53.1285 must indicate that the application includes this notification.
    (b) Environmental report. (1) The application must contain an 
environmental report as required by Sec.  51.54 of this chapter.
    (2) If the ML application references a standard design 
certification, the environmental report need not contain a discussion 
of severe accident mitigation design alternatives for the manufactured 
reactor as used in a commercial nuclear plant.
    (c) Safeguards information. The application must contain a 
description of the program to protect safeguards information against 
unauthorized disclosure in accordance with the requirements in 
Sec. Sec.  73.21 and 73.22 of this chapter, as applicable.
    (d) Performance demonstration. A description of how the performance 
of each design feature has been demonstrated capable of fulfilling 
functional design criteria considering interdependent effects through 
either analysis, appropriate test programs, prototype testing, 
operating experience,

[[Page 87091]]

or a combination thereof, in accordance with Sec.  53.440(a).


Sec.  53.1285  Review of applications.

    (a) Standards for review of applications. Applications for MLs 
under this part will be reviewed according to the applicable standards 
set out in this subpart as well as applicable standards in this part 
and 10 CFR parts 20, 25, 26, 51, 70, 71, 73, and 75.
    (b) Administrative review of applications, hearings. A proceeding 
on an ML is subject to all applicable procedural requirements contained 
in 10 CFR part 2, including the requirements for docketing in Sec.  
2.101(a)(1) through (4) of this chapter, and the requirements for 
issuance of a notice of proposed action in Sec.  2.105 of this chapter, 
provided, however, that the designated sections may not be construed to 
require that the environmental report or draft or final environmental 
impact statement include an assessment of the benefits of constructing 
and/or operating the manufactured reactor or an evaluation of 
alternative energy sources. All hearings on MLs are governed by the 
hearing procedures contained in 10 CFR part 2, subparts C, E, G, L, and 
N.


Sec.  53.1286  Referral to the Advisory Committee on Reactor 
Safeguards.

    The Commission must refer a copy of the application to the ACRS. 
The ACRS must report on those portions of the application which concern 
safety.


Sec.  53.1287  Issuance of manufacturing licenses.

    (a) After completing any hearing under Sec.  53.1285(b), and 
receiving the report submitted by the ACRS, the Commission may issue an 
ML if the Commission finds that--
    (1) Applicable standards and requirements of the Act and the 
Commission's regulations have been met;
    (2) There is reasonable assurance that the manufactured reactor 
will be manufactured, and can be transported, incorporated into a 
commercial nuclear plant, and operated in conformity with the ML, the 
provision of the Act, and the Commission's regulations;
    (3) The proposed manufactured reactor can be incorporated into a 
commercial nuclear plant and operated at sites having characteristics 
that fall within the site parameters postulated for the design of the 
manufactured reactors without undue risk to the health and safety of 
the public;
    (4) The applicant is technically qualified to design and 
manufacture the proposed manufactured reactor;
    (5) The proposed ITAAC are necessary and sufficient, within the 
scope of the ML, to provide reasonable assurance that the manufactured 
reactor has been manufactured and will be operated in conformity with 
the license, the provisions of the Act, and the Commission's 
regulations;
    (6) The issuance of a license to the applicant will not be inimical 
to the common defense and security or to the health and safety of the 
public; and
    (7) The findings required by subpart A of 10 CFR part 51 have been 
made.
    (b) Each ML issued under this subpart must specify--
    (1) Terms and conditions as the Commission deems necessary and 
appropriate;
    (2) Technical specifications for operation of the manufactured 
reactor, as the Commission deems necessary and appropriate;
    (3) Site parameters and design characteristics for the manufactured 
reactor;
    (4) The interface requirements to be met by the site-specific 
elements of the facility, such as the energy conversions systems and 
ultimate heat sink, not within the scope of the manufactured reactor; 
and
    (5) The entity with design authority for the manufactured reactor 
covered by the license.


Sec.  53.1288  Finality of manufacturing licenses.

    (a)(1) Notwithstanding any provision in Sec.  53.1590, during the 
term of an ML issued under this part the Commission may not modify, 
rescind, or impose new requirements on the design of the manufactured 
reactor, or the requirements for the manufacture of the manufactured 
reactor, unless the Commission determines that a modification is 
necessary to bring the design of the reactor or its manufacture into 
compliance with the Commission's requirements applicable and in effect 
at the time the ML was issued, or to provide reasonable assurance of 
adequate protection to public health and safety or common defense and 
security.
    (2) Any modification to the design of a manufactured reactor that 
is imposed by the Commission under paragraph (a)(1) of this section 
will be applied to all manufactured reactors manufactured under the 
license, including those that have already been transported and sited, 
except those manufactured reactors to which the modification has been 
rendered technically irrelevant by action taken under Sec.  53.1530 or 
paragraph (b) of this section.
    (3) In making the findings required under this part for issuance of 
a COL, in any hearing under Sec.  53.1452, or in any enforcement 
hearing other than one initiated by the Commission under paragraph 
(a)(1) of this section, for which a manufactured reactor manufactured 
under this subpart is referenced or used, the Commission must treat as 
resolved those matters resolved in the proceeding on the application 
for issuance or renewal of the ML, including the adequacy of design of 
the manufactured reactor, the costs and benefits of severe accident 
mitigation design alternatives, and the bases for not incorporating 
severe accident mitigation design alternatives into the design of the 
manufactured reactor to be manufactured.
    (b) An applicant who references or uses a manufactured reactor 
manufactured under an ML under this part may include in the application 
a request for a departure from the design characteristics, site 
parameters, terms and conditions, or approved design of the 
manufactured reactor. The Commission may grant a request only if it 
determines that the departure will comply with the requirements of 
Sec.  53.080, and that the special circumstances outweigh any decrease 
in safety that may result from the reduction in standardization caused 
by the departure. The granting of a departure on request of an 
applicant is subject to litigation in the same manner as other issues 
in the COL hearing.


Sec.  53.1291  Duration of manufacturing licenses.

    An ML issued under this part is valid for not less than 5, nor more 
than 15 years from the date of issuance. Upon expiration of the ML, the 
manufacture of any uncompleted manufactured reactors must cease unless 
a timely application for renewal has been docketed with the NRC.


Sec.  53.1293  Transfer of manufacturing licenses.

    An ML may be transferred under Sec.  53.1570.


Sec.  53.1295  Renewal of manufacturing licenses.

    (a)(1) Not less than 12 months, nor more than 5 years before the 
expiration of the ML, or any later renewal period, the holder of the ML 
issued under this part may apply for a renewal of the license. An 
application for renewal must contain all information necessary to bring 
up to date the information and data contained in the previous 
application.
    (2) The filing of an application for a renewed license must be in 
accordance with subpart A of 10 CFR part 2 and Sec.  53.1100.

[[Page 87092]]

    (3) An ML issued under this part, either original or renewed, for 
which a timely application for renewal has been filed, remains in 
effect until the Commission has made a final determination on the 
renewal application, provided, however, that the holder of an ML may 
not begin manufacture of a manufactured reactor less than 6 months 
before the expiration of the license.
    (4) Any person whose interest may be affected by renewal of the 
license may request a hearing on the application for renewal. The 
request for a hearing must comply with Sec.  2.309 of this chapter. If 
a hearing is granted, notice of the hearing will be published in 
accordance with Sec.  2.104 of this chapter.
    (5) The Commission must refer a copy of the application for renewal 
to the ACRS. The ACRS must report on those portions of the application 
which concern safety.
    (b) The Commission may grant the renewal if the Commission 
determines--
    (1) The ML complies with the Act and the Commission's regulations 
and orders applicable and in effect at the time the ML was originally 
issued; and
    (2) Any new requirements the Commission may wish to impose are--
    (i) Necessary for adequate protection to public health and safety 
or common defense and security;
    (ii) Necessary for compliance with the Commission's regulations and 
orders applicable and in effect at the time the ML was originally 
issued; or
    (iii) A substantial increase in overall protection of the public 
health and safety or the common defense and security to be derived from 
the new requirements, and the direct and indirect costs of 
implementation of those requirements are justified in view of this 
increased protection.
    (c) A renewed ML may be issued for a term of not less than 5, nor 
more than 15 years, plus any remaining years on the ML then in effect 
before renewal. The renewed license must be subject to the requirements 
of Sec.  53.1288.


Sec.  53.1300  Construction permits.

    Sections 53.1300 through 53.1348 set out the requirements and 
procedures applicable to Commission issuance of a CP for commercial 
nuclear plants. A CP for the construction of a commercial nuclear plant 
under this part will be issued before the issuance of an OL if the 
application is otherwise acceptable and will be converted upon 
completion of the facility and Commission action, into an OL as 
provided under Sec. Sec.  53.1360 through 53.1405.


Sec.  53.1306  Contents of applications for construction permits; 
general information.

    An application for a CP must include the information required by 
Sec.  53.1109 and the following information:
    (a) Information sufficient to demonstrate to the Commission the 
financial qualification of the applicant to carry out, under the 
regulations in this chapter, the activities for which the permit is 
sought. As applicable, the following should be provided:
    (1) The information that demonstrates that the applicant possesses 
or has reasonable assurance of obtaining the funds necessary to cover 
estimated construction costs and related fuel cycle costs, including 
estimates of the total construction costs and related fuel cycle costs 
of the facility and must indicate the source(s) of funds to cover these 
costs.
    (2) Each application for a CP submitted by a newly formed entity 
organized for the primary purpose of constructing and operating a 
facility must also include information showing:
    (i) The legal and financial relationships the entity has or 
proposes to have with its stockholders or owners;
    (ii) The stockholders' or owners' financial ability to meet any 
contractual obligation to the entity that they have incurred or 
proposed to incur; and
    (iii) Any other information considered necessary by the Commission 
to enable it to determine the applicant's financial qualification; and
    (3) The Commission may request an established entity or newly-
formed entity to submit additional or more detailed information 
respecting its financial arrangements and status of funds if the 
Commission considers this information appropriate. This may include 
information regarding an applicant's ability to continue the conduct of 
the activities authorized by the CP and to decommission the facility.
    (b) If the applicant proposes to construct or alter a facility, the 
application must state the earliest and latest dates for completion of 
the construction or alteration.


Sec.  53.1309  Contents of applications for construction permits; 
technical information.

    The application must contain a Preliminary Safety Analysis Report 
(PSAR) that describes the facility and the limits on its operation and 
presents a preliminary safety analysis of the SSCs of the facility as a 
whole. The PSAR must include the following information, at a level of 
detail sufficient to enable the Commission to reach a conclusion on 
safety matters that must be resolved by the Commission before issuance 
of a CP:
    (a)(1) Site information. An application for a CP for a commercial 
nuclear reactor must include the site information equivalent to that 
required for an early site permit in Sec.  53.1146(a)(1)(iv) through 
(x).
    (2) Design information. Except as specified in this paragraph, an 
application for a CP for a commercial nuclear plant must include the 
design information equivalent to that required for a standard design 
certification as defined in Sec.  53.1239(a)(2) through (27).
    (i) Quality assurance program. A description of the QAP to be 
applied to the design, fabrication, construction, and testing of the 
SSCs of the facility under Sec.  53.610(a)(6), including a discussion 
of how the requirements of appendix B to part 50 of this chapter will 
be satisfied.
    (ii) Preliminary design information. The information provided in 
the application may include some aspects of the design that are not 
fully developed, and the information is therefore preliminary. The 
completed design, including any changes during construction, must be 
described in the FSAR required in Sec.  53.1369 that supports an 
application for an OL.
    (iii) Planned research or testing. Descriptions of how design 
features and related functional design criteria will fulfill the safety 
criteria in subpart B, or more restrictive alternative criteria adopted 
under Sec.  53.470, and how that has been or will be demonstrated 
through either analysis, appropriate test programs, experience, or a 
combination thereof. Where any design feature has not been fully 
developed or demonstrated to fulfill the functional design criteria at 
the time of an application for a CP, the applicant must provide a plan 
for future analysis, research and development, test programs, gathering 
of experience, or a combination thereof to provide reasonable 
confidence that the required demonstration will be available for an 
application for an OL.
    (iv) Programmatic controls. Descriptions of the programmatic 
controls may include those to be provided in the FSAR or other 
licensing basis documents because they are necessary to achieve and 
maintain the reliability and capability of SSCs relied upon to 
demonstrate compliance with the established safety criteria and 
functional design criteria required in subpart B, and to maintain 
consistency with analyses required by Sec.  53.450.
    (3) Technical qualifications. A description of the technical 
qualifications of the applicant to engage in the proposed activities 
under the regulations in this chapter.

[[Page 87093]]

    (4) Emergency preparedness. A description of the applicant's 
preliminary plans for coping with emergencies based on:
    (i) Except as provided in paragraph (a)(4)(ii) of this section, the 
requirements in appendix E to part 50.
    (ii) For a commercial nuclear plant consisting of either small 
modular reactors or non-light-water reactors, the requirements in 
either Sec.  50.160 or appendix E to part 50.
    (5) Physical security. A report that provides a preliminary 
description of how the site characteristics support the development of 
adequate security plans and measures consistent with the requirements 
in Sec.  53.540.
    (6) Fitness-for-duty program. A description of the fitness-for-duty 
(FFD) program required by 10 CFR part 26 and its implementation.
    (b) A description of the program to protect Safeguards Information 
against unauthorized disclosure in accordance with the requirements in 
Sec. Sec.  73.21 and 73.22 of this chapter, as applicable.


Sec.  53.1312  Contents of applications for construction permits; other 
application content.

    (a) In addition to the PSAR, the application must include the 
following:
    (1) An environmental report either under Sec.  51.50(a) of this 
chapter if an LWA under Sec.  53.1130 is not requested in conjunction 
with the CP application, or under Sec. Sec.  51.49 and 51.50(a) of this 
chapter if an LWA is requested in conjunction with the CP application; 
or
    (2) If the applicant wishes to request that an LWA under Sec.  
53.1130 be issued before issuance of the CP, the information otherwise 
required by Sec.  53.1130, in accordance with either Sec.  2.101(a)(1) 
through (a)(5), or Sec.  2.101(a)(9) of this chapter.
    (b) If the CP application references an early site permit, standard 
design approval, or standard design certification issued under this 
part, then the following requirements apply:
    (1) The PSAR need not contain information or analyses submitted to 
the Commission in connection with the referenced NRC approval, permit, 
or certification, provided, however, that the PSAR incorporates the 
material by reference and confirms that the site and design of the 
facility falls within parameter values postulated in the referenced NRC 
approval, permit, or certification.
    (2) The PSAR must provide a means to demonstrate that all terms and 
conditions that have been included in the referenced NRC approval, 
permit, or certification will be satisfied by the date of issuance of 
the OL, as appropriate. If the PSAR does not demonstrate that each site 
characteristic falls within the corresponding postulated site parameter 
and each design characteristic of the facility falls within the 
corresponding postulated design parameter, the application must justify 
a departure, variance, or exemption from the referenced NRC approval, 
license, or certification in regard to that particular site or design 
characteristic in compliance with the requirements of this part.
    (3) If a referenced early site permit approves complete and 
integrated emergency plans, or major features of emergency plans, then 
the PSAR must include any new or additional information that updates 
and corrects the information that was provided under Sec.  
53.1146(b)(2) and discuss whether the new or additional information 
materially changes the bases for compliance with the applicable 
requirements.


Sec.  53.1315  Review of applications.

    (a) Standards for review of applications. Applications filed under 
this part will be reviewed according to the standards set out in this 
part and 10 CFR parts 20, 51, 73, and 140.
    (b) Administrative review of applications; hearings. A proceeding 
on a CP application is subject to all applicable procedural 
requirements contained in 10 CFR part 2, including the requirements for 
docketing (Sec.  2.101 of this chapter) and issuance of a notice of 
hearing (Sec.  2.104 of this chapter). All hearings on CP applications 
are governed by the procedures contained in 10 CFR part 2.


Sec.  53.1318  Finality of referenced NRC approvals, permits, and 
certifications.

    If the application for a CP under this part references an early 
site permit, standard design approval, or standard design 
certification, the scope and nature of matters resolved for the 
application are governed by the relevant provisions addressing 
finality, including Sec. Sec.  53.1188, 53.1221, and 53.1263.


Sec.  53.1324  Referral to the Advisory Committee on Reactor 
Safeguards.

    The Commission must refer a copy of the application to the ACRS. 
The ACRS must report on those portions of the application that concern 
safety and must apply the standards referenced in Sec.  53.1315, in 
accordance with the finality provisions in Sec.  53.1318.


Sec.  53.1327  Authorization to conduct limited work authorization 
activities.

    (a) If the application does not reference an early site permit 
which authorizes the holder to perform the activities under Sec.  
53.1130, the applicant may not perform those activities without 
obtaining the separate authorization required by Sec.  53.1130. 
Authorization may be granted only after the presiding officer in the 
proceeding on the application has made the findings and determination 
required by Sec.  53.1130(b)(1)(ii) and (iv), and the Director, Office 
of Nuclear Reactor Regulation makes the determination required by Sec.  
53.1130(b)(1)(iii).
    (b) If, after an applicant has performed the activities permitted 
by paragraph (a) of this section, the application for the CP is 
withdrawn or denied, then the applicant must implement an approved site 
redress plan.


Sec.  53.1330  Exemptions, departures, and variances.

    (a) Applicants for a CP under this part, or any amendment to a CP, 
may include in the application a request for an exemption from one or 
more of the Commission's regulations. The Commission may grant a 
request if it determines that the exemption complies with Sec.  53.080.
    (b) An applicant for a CP who has filed an application referencing 
an NRC approval, permit, or certification issued under this part may 
include in the application a request for exemptions, departures, or 
variances related to the subject referenced NRC approval, permit, or 
certification. In determining whether to grant the departure, variance, 
or exemption, the Commission must apply the same technically relevant 
criteria as were applicable to the application for the original or 
renewed approval, license, or certification.


Sec.  53.1333  Issuance of construction permits.

    (a) After conducting a hearing in accordance with Sec.  53.1315 and 
receiving the report submitted by the ACRS, the Commission may issue a 
CP only if the Commission finds that--
    (1) The applicant has described the proposed design of the facility 
and has identified the major features or components incorporated 
therein for the protection of the health and safety of the public;
    (2) Such further technical or design information as may be required 
to complete the safety analysis, and which can reasonably be left for 
later consideration, will be supplied in the FSAR;
    (3) Safety features or components, if any, that require research 
and development have been described by the applicant and the applicant 
has identified, and there will be conducted,

[[Page 87094]]

a research and development program reasonably designed to resolve any 
safety questions associated with such features or components; and
    (4) On the basis of the foregoing, there is reasonable assurance of 
the following--
    (i) Such safety questions will be satisfactorily resolved at or 
before the latest date stated in the application for completion of 
construction of the proposed facility; and
    (ii) Taking into consideration the site criteria contained subpart 
D to this part, the proposed facility can be constructed and operated 
at the proposed location without undue risk to the health and safety of 
the public.
    (b) A CP must contain the terms and conditions for the permit, as 
the Commission deems necessary and appropriate. The Commission may, in 
its discretion, incorporate in any CP provisions requiring the 
applicant to furnish periodic reports of the progress and results of 
research and development programs designed to resolve safety questions.


Sec.  53.1336  Finality of construction permits.

    Notwithstanding any provision in Sec.  53.1590, a CP constitutes an 
authorization to proceed with construction but does not constitute 
Commission approval of the safety of any design feature or 
specification unless the applicant specifically requests such approval 
and such approval is incorporated in the permit. The applicant, at its 
option, may request such approvals in the CP or by amendment to the CP. 
If approved by the NRC and included in the permit, the NRC will 
consider modifications to the approved design features or 
specifications in accordance with Sec.  53.1590.


Sec.  53.1342  Duration of construction permits.

    (a) A CP will state the earliest and latest dates for completion of 
construction or alteration of the facility, not to exceed 40 years from 
date of issuance.
    (b) If the proposed construction or alteration of the facility is 
not completed by the latest completion date, the CP shall expire, and 
all rights are forfeited. However, upon good cause shown, the 
Commission will extend the completion date for a reasonable period of 
time. The Commission will recognize, among other things, developmental 
problems attributed to the experimental nature of the facility or fire, 
flood explosion, strike, sabotage, domestic violence, enemy action an 
act of the elements, and other acts beyond the control of the permit 
holder, as a basis for extending the completion date.


Sec.  53.1345  Transfer of construction permits.

    A CP may be transferred under Sec.  53.1570.


Sec.  53.1348  Termination of construction permits.

    When a permit holder has determined to permanently cease 
construction, the holder must, within 30 days, submit a written 
certification to the NRC.


Sec.  53.1360  Operating licenses.

    Sections 53.1360 through 53.1405 set out the requirements and 
procedures applicable to Commission issuance of an OL for a nuclear 
power facility.


Sec.  53.1366  Contents of applications for operating licenses; general 
information.

    An application for an OL must include the information required by 
Sec.  53.1109 and the following information:
    (a) Except for an electric utility applicant, information 
sufficient to demonstrate to the Commission the financial qualification 
of the applicant to carry out, in accordance with the regulations in 
this chapter, the activities for which the license is sought. As 
applicable, the following should be provided:
    (1) The applicant must submit information that demonstrates the 
applicant possesses or has reasonable assurance of obtaining the funds 
necessary to cover estimated operation costs for the period of the 
license. The applicant must submit estimates for total annual operating 
costs for each of the first 5 years of operation of the facility. The 
applicant must also indicate the source(s) of funds to cover these 
costs.
    (2) Each application for an OL submitted by a newly-formed entity 
organized for the primary purpose of operating the facility must also 
include information showing--
    (i) The legal and financial relationships the entity has or 
proposes to have with its stockholders or owners;
    (ii) The stockholders' or owners' financial ability to meet any 
contractual obligation to the entity which they have incurred or 
proposed to incur; and
    (iii) Any other information considered necessary by the Commission 
to enable it to determine the applicant's financial qualification.
    (3) The Commission may request an established entity or newly 
formed entity to submit additional or more detailed information 
respecting its financial arrangements and status of funds if the 
Commission considers this information appropriate. This may include 
information regarding a licensee's ability to continue the conduct of 
the activities authorized by the license and to decommission the 
facility.
    (b) The application must include information in the form of a 
report, as described in subpart G, indicating how reasonable assurance 
will be provided that funds will be available to decommission the 
facility, including a copy of the financial instrument obtained to 
satisfy the requirements of Sec.  53.1040.


Sec.  53.1369  Contents of applications for operating licenses; 
technical information.

    Final Safety Analysis Report. The application must contain an FSAR 
that describes the facility and the limits on its operation and 
presents a safety analysis of the SSCs of the facility as a whole. The 
FSAR must include the following information, at a level of detail 
sufficient to enable the Commission to reach a final conclusion on all 
safety matters that must be resolved by the Commission before issuance 
of an OL. The FSAR must include the following information:
    (a) Site information. An application for an OL for a commercial 
nuclear reactor must include the site information equivalent to that 
required for an early site permit in Sec.  53.1146(a)(1)(iv) through 
(x), including all current information, such as the results of 
environmental and meteorological monitoring programs, which has been 
developed since issuance of the CP, relating to site evaluation factors 
identified in this part.
    (b) Design information. Except as specified in this paragraph, an 
FSAR for an OL for a commercial nuclear plant must include the final 
design information equivalent to that required for a standard design 
certification as defined in Sec.  53.1239(a)(2) through (7), (a)(9), 
and (a)(11) through (a)(27).
    (1) The completed design, including any changes during 
construction, must be described.
    (2) Where any design feature had not been fully developed or 
demonstrated at the time of application for the CP, the applicant must 
provide the analysis, research and development, test programs, 
gathering of experience, or a combination thereof to provide the 
required demonstration to fulfill the functional design criteria.
    (c) Technical qualifications. A description of the technical 
qualifications of the applicant to engage in the proposed activities in 
accordance with the regulations in this chapter.

[[Page 87095]]

    (d) Integrity assessment program. A description of an Integrity 
Assessment Program that addresses the elements described in Sec.  
53.870.
    (e) Safeguards information. A description of the program to protect 
Safeguards Information against unauthorized disclosure in accordance 
with the requirements in Sec. Sec.  73.21 and 73.22 of this chapter, as 
applicable.
    (f) Emergency response facility or facilities. Description of 
location and capabilities to be established for command and control, 
support, and coordination of onsite and offsite, as applicable, 
functions during reactor accident conditions.
    (g) Role of personnel. (1) A description of the completed 
assessments related to the role of personnel in ensuring safe 
operations considering the analyses required by Sec.  53.730. These 
assessments must include the following:
    (i) Human factors engineering design requirements of Sec.  
53.730(a);
    (ii) Human system interface design requirements of Sec.  53.730(b);
    (iii) Concept of operations of Sec.  53.730(c);
    (iv) Functional requirements analysis and function allocation of 
Sec.  53.730(d);
    (2) A description of the program to be used for evaluating and 
applying operating experience as required by Sec.  53.730(e);
    (3) A staffing plan and supporting analyses as required by Sec.  
53.730(f).
    (h) Training, examination, and proficiency programs. (1) A 
description of the training, examination, and proficiency programs 
required by Sec.  53.730(g);
    (2) A description of the training programs required by Sec.  
53.830.
    (i) Emergency plan. Emergency plans complying with the requirements 
of Sec.  53.855.
    (1) Include all emergency plan certifications, as applicable, that 
have been obtained from the State, local, and participating Tribal 
governmental agencies with emergency planning responsibilities that are 
wholly or partially within the EPZ plume exposure pathway. These 
certifications must state that--
    (i) The proposed emergency plans are practicable;
    (ii) These agencies are committed to participating in any further 
development of the plans, including any required field demonstrations; 
and
    (iii) These agencies are committed to executing their 
responsibilities under the plans in the event of an emergency.
    (2) If certifications cannot be obtained after sustained, good 
faith efforts by the applicant, then the application must contain 
information, including a utility plan, sufficient to show that the 
proposed plans provide reasonable assurance that adequate protective 
measures can and will be taken in the event of a radiological emergency 
at the site.
    (3) If complete and integrated emergency plans were approved as 
part of an early site permit, or submitted, reviewed, and approved as 
part of the CP application, new certifications that demonstrate 
compliance with the requirements of paragraph (i)(1) of this section 
are not required.
    (j) Organization. A description of the applicant's organizational 
structure, allocations of responsibilities and authorities, and 
personnel qualifications requirements for operation.
    (k) Maintenance program. A description of a maintenance program 
under Sec.  53.715.
    (l) Quality assurance. A description of the QAP that demonstrates 
compliance with the requirements under Sec.  53.865.
    (m) Radiation protection program. A radiation protection program 
description under Sec.  53.850.
    (n) Security program. A physical security plan that describes how 
the applicant will comply with Sec.  53.860 (and 10 CFR part 11, if 
applicable, including the identification and description of jobs as 
required by Sec.  11.11(a) of this chapter, at the proposed facility). 
The plan must list tests, inspections, audits, and other means to be 
used to demonstrate compliance with the requirements of 10 CFR parts 11 
and 73, if applicable.
    (o) Safeguards contingency plan. A safeguards contingency plan in 
accordance with the criteria set forth in appendix C to 10 CFR part 73. 
The safeguards contingency plan must include plans for dealing with 
threats, thefts, and radiological sabotage, as defined in 10 CFR part 
73, relating to the SNM and nuclear facilities licensed under this 
chapter and in the applicant's possession and control. Each application 
for this type of license must include the information contained in the 
applicant's safeguards contingency plan. (Implementing procedures 
required for this plan need not be submitted for approval.) \1\
    (p) Security training and qualification. A training and 
qualification plan that describes how the applicant will demonstrate 
compliance with the criteria set forth in Sec.  73.100 of this chapter 
or appendix B to 10 CFR part 73.
    (q) Cybersecurity plan. A cybersecurity plan in accordance with the 
criteria set forth in Sec.  73.54 or Sec.  73.110 of this chapter.
    (r) Security, safeguards and cybersecurity plan implementation. A 
description of the implementation of the physical security plan, 
safeguards contingency plan, training and qualification plan, and 
cybersecurity plan. Each applicant who prepares a physical security 
plan, a safeguards contingency plan, a training and qualification plan, 
or a cybersecurity plan must protect the plans and other related 
Safeguards Information against unauthorized disclosure in accordance 
with the requirements of Sec. Sec.  73.21 and 73.22 of this chapter.
    (s) Fire protection program. A description of the fire protection 
program under Sec.  53.875.
    (t) Inservice inspection/inservice testing program. A description 
of the inservice inspection and inservice testing programs under Sec.  
53.880.
    (u) [Reserved]
    (v) [Reserved]
    (w) General employee training. A description of the training 
program required to demonstrate compliance with Sec.  53.830 and its 
implementation.
    (x) Fitness-for-duty program. A description of the FFD program 
required by 10 CFR part 26 and its implementation.
    (y) Other programs. A description and evaluation of the results of 
the applicant's programs, including research and development, if any, 
to demonstrate that any safety questions identified at the CP stage 
have been resolved.
    (z) Safety design feature performance. A description of how the 
performance of each safety design feature has been demonstrated capable 
of fulfilling functional design criteria considering interdependent 
effects through either analysis, appropriate test programs, prototype 
testing, operating experience, or a combination thereof, in accordance 
with Sec.  53.440(a).
    (aa) Technical specifications. Proposed technical specifications 
prepared in accordance with the requirements of Sec.  53.710(a).

    \1\ A physical security plan that contains all the information 
required in both Sec.  73.55 or Sec.  73.100 of this chapter and 
appendix C to 10 CFR part 73 satisfies the requirement for a 
contingency plan.


Sec.  53.1372  Contents of applications for operating licenses; other 
application content.

    In addition to the FSAR, the application must also include the 
following:
    (a) Environmental report. An environmental report in accordance 
with Sec.  51.53(b) of this chapter.
    (b) Availability controls (if not included in the FSAR). A 
description of

[[Page 87096]]

the controls on plant operations, including availability controls, to 
provide reasonable confidence of safe operation and that the 
configurations and special treatments for SR and NSRSS SSCs provide the 
capabilities and reliabilities required to satisfy the safety criteria 
of Sec.  53.220, or more restrictive alternative criteria adopted under 
Sec.  53.470, if not addressed by Technical Specifications under Sec.  
53.1369(aa).


Sec.  53.1375  Review of applications.

    (a) Standards for review of applications. Applications filed under 
this part will be reviewed according to the standards set out in 10 CFR 
parts 20, 26, 51, 53, 73, and 140.
    (b) Administrative review of applications; hearings. A proceeding 
on an OL is subject to all applicable procedural requirements contained 
in 10 CFR part 2, including the requirements for docketing (Sec.  2.101 
of this chapter) and issuance of a notice of hearing (Sec.  2.104 of 
this chapter). All hearings on OLs are governed by the procedures 
contained in 10 CFR part 2.


Sec.  53.1381  Referral to the Advisory Committee on Reactor 
Safeguards.

    The Commission must refer a copy of the application to the ACRS. 
The ACRS must report on those portions of the application that concern 
safety and must apply the standards referenced in Sec.  53.1375.


Sec.  53.1384  Exemptions, departures, and variances.

    (a) Applicants for an OL under this part, or any amendment to an 
OL, may include in the application a request for an exemption from one 
or more of the Commission's regulations. The Commission may grant an 
exemption request if it determines that the exemption complies with 
Sec.  53.080.
    (b) An applicant for an OL who has filed an application referencing 
an NRC approval, permit, license, or certification issued under this 
part may include in the application a request for departures, 
variances, or exemptions related to the subject referenced NRC 
approval, permit, license, or certification. In determining whether to 
grant the departure, variance, or exemption, the Commission must apply 
the same technically relevant criteria as were applicable to the 
application for the original or renewed approval, license, or 
certification.


Sec.  53.1387  Issuance of operating licenses.

    Upon completion of the construction or alteration of a facility, in 
compliance with the terms and conditions of the construction permit and 
subject to any necessary testing of the facility for health or safety 
purposes, the Commission will, in the absence of good cause shown to 
the contrary, issue an OL or an appropriate amendment of the license, 
as the case may be.
    (a)(1) After receiving the report submitted by the ACRS, the 
Commission may issue an OL if the Commission finds that--
    (i) Construction of the facility has been substantially completed 
in conformity with the CP and the application as amended, the 
provisions of the Act, and the rules and regulations of the Commission;
    (ii) Any required notifications to other agencies or bodies have 
been duly made;
    (iii) The facility will operate in conformity with the application 
as amended, the provisions of the Act, and the rules and regulations of 
the Commission;
    (iv) There is reasonable assurance that--
    (A) the activities authorized by the OL can be conducted without 
endangering the health and safety of the public; and
    (B) such activities will be conducted in compliance with the 
regulations in this chapter.
    (v) The applicant is technically and financially qualified to 
engage in the activities authorized, however, no finding of financial 
qualification is necessary for an electric utility applicant for an OL;
    (vi) Issuance of the license will not be inimical to the common 
defense and security or to the health and safety of the public;
    (vii) The applicable provisions of 10 CFR part 140 have been 
satisfied; and
    (viii) The findings required by subpart A of 10 CFR part 51 have 
been made.
    (2) [Reserved]
    (b) [Reserved]
    (c) The OL will include appropriate provisions with respect to any 
uncompleted items of construction and such limitations or conditions as 
are required to assure that operation during the period of the 
completion of such items will not endanger public health and safety.
    (d) The Commission will issue an OL in such form and containing 
such conditions and limitations, including technical specifications, as 
it deems necessary and appropriate.


Sec.  53.1390  Backfitting of operating licenses.

    After issuance of an OL, the Commission may not modify, add, or 
delete any term or condition of the OL, except in accordance with the 
provisions of Sec.  53.1590.


Sec.  53.1396  Duration of operating licenses.

    The Commission will issue an OL under this part for the term 
requested by the applicant, not to exceed 40 years from the date of 
issuance, or for the estimated useful life of the facility if the 
Commission determines that the estimated useful life is less than the 
term requested.


Sec.  53.1399  Transfer of an operating license.

    An OL may be transferred under Sec.  53.1570.


Sec.  53.1402  Application for renewal.

    The filing of an application for a renewed license must be in 
accordance with Sec.  53.1595.


Sec.  53.1405  Continuation of an operating license.

    Each OL for a facility that has permanently ceased operations 
continues in effect beyond the expiration date to authorize ownership 
and possession of the facility until the Commission notifies the 
licensee in writing that the license is terminated. During this period 
of continued effectiveness, the licensee must--
    (a) Take actions necessary to decommission and decontaminate the 
facility and continue to maintain the facility, including, where 
applicable, the storage, control, and maintenance of the spent fuel in 
a safe condition; and
    (b) Conduct activities in accordance with all other restrictions 
applicable to the facility in accordance with the NRC's regulations and 
the provisions of the OL for the facility.


Sec.  53.1410  Combined licenses.

    Sections 53.1410 through 53.1461 set out the requirements and 
procedures applicable to Commission issuance of COLs for commercial 
nuclear plants under this part.


Sec.  53.1413  Contents of applications for combined licenses; general 
information.

    An application for a COL must include the information required by 
Sec.  53.1109 and the following information:
    (a) Except for an electric utility applicant in regard to financial 
assurance required after a Commission finding under Sec.  53.1452, the 
application must include information sufficient to demonstrate to the 
Commission the financial qualification of the applicant to carry out, 
in accordance with the regulations in this chapter, the activities for 
which the permit or license is sought. As applicable, the following 
should be provided:
    (1) The applicant must submit information that demonstrates that 
the applicant possesses or has reasonable

[[Page 87097]]

assurance of obtaining the funds necessary to cover estimated 
construction costs and related fuel cycle costs. The applicant must 
submit estimates of the total construction costs of the facility and 
related fuel cycle costs and must indicate the source(s) of funds to 
cover these costs.
    (2) The applicant must submit information that demonstrates the 
applicant possesses or has reasonable assurance of obtaining the funds 
necessary to cover estimated operation costs for the period of the 
license. The applicant must submit estimates for total annual operating 
costs for each of the first 5 years of operation of the facility. The 
applicant must also indicate the source(s) of funds to cover these 
costs.
    (3) Each application for a COL submitted by a newly-formed entity 
organized for the primary purpose of constructing and operating a 
facility must also include information showing--
    (i) The legal and financial relationships the entity has or 
proposes to have with its stockholders or owners; and
    (ii) The stockholders' or owners' financial ability to meet any 
contractual obligation to the entity which they have incurred or 
proposed to incur; and
    (iii) Any other information considered necessary by the Commission 
to enable it to determine the applicant's financial qualification.
    (4) The Commission may request an established entity or newly 
formed entity to submit additional or more detailed information 
respecting its financial arrangements and status of funds if the 
Commission considers this information appropriate. This may include 
information regarding a licensee's ability to continue the conduct of 
the activities authorized by the license and to decommission the 
facility.
    (b) The application must include information in the form of a 
report, as described in subpart G of this part, indicating how 
reasonable assurance will be provided that funds will be available to 
decommission the facility.


Sec.  53.1416  Contents of applications for combined licenses; 
technical information.

    (a) Final Safety Analysis Report. The application must contain an 
FSAR that describes the facility and the limits on its operation and 
presents a safety analysis of the SSCs of the facility as a whole. The 
Commission will require, before issuance of a COL, that engineering 
documents, such as analyses, drawings, procurement specifications, or 
construction and installation specifications, be completed and 
available for audit if the more detailed information is necessary for 
the Commission to verify the information in the application and make 
its safety determination. The FSAR must include the following 
information, at a level of detail sufficient to enable the Commission 
to reach a final conclusion on all safety matters that must be resolved 
by the Commission before issuance of a COL:
    (1) Site information. An application for a COL for a commercial 
nuclear reactor must include the site information required for an early 
site permit in Sec.  53.1146(a)(1)(iv) through (x).
    (2) Design information. An application for a COL for a commercial 
nuclear plant must include the design information equivalent to that 
required for a standard design certification as defined in Sec.  
53.1239(a)(2) through (7), (a)(9), and (a)(11) through (27).
    (3) Technical qualifications. A description of the technical 
qualifications of the applicant to engage in the proposed activities in 
accordance with the regulations in this chapter.
    (4) Integrity assessment program. A description of an Integrity 
Assessment Program that addresses the elements described in Sec.  
53.870.
    (5) Safeguards information. A description of the program to protect 
Safeguards Information against unauthorized disclosure in accordance 
with the requirements in Sec. Sec.  73.21 and 73.22 of this chapter, as 
applicable.
    (6) Emergency response facility or facilities. Description of the 
locations and capabilities to be established for command and control, 
support, and coordination of onsite and offsite, as applicable, 
functions during reactor accident conditions.
    (7) Role of personnel. (i) A description of the completed 
assessments related to the role of personnel in ensuring safe 
operations considering the analyses required by Sec.  53.730. These 
assessments must include the following:
    (A) Human factors engineering design requirements of Sec.  
53.730(a);
    (B) Human system interface design requirements of Sec.  53.730(b);
    (C) Concept of operations of Sec.  53.730(c); and
    (D) Functional requirements analysis and function allocation of 
Sec.  53.730(d);
    (ii) A description of the program to be used for evaluating and 
applying operating experience as required by Sec.  53.730(e);
    (iii) A staffing plan and supporting analyses as required by Sec.  
53.730(f).
    (8) Training, examination, and proficiency programs. (i) A 
description of the training, examination, and proficiency programs 
required by Sec.  53.730(g); and
    (ii) A description of the training programs required by Sec.  
53.830.
    (9) Emergency plan. Emergency plans complying with the requirements 
of Sec.  53.855.
    (i) The emergency plan must include, as applicable, all emergency 
plan certifications that have been obtained from the State, local, and 
participating Tribal governmental agencies with emergency planning 
responsibilities. The certifications must state that--
    (A) The proposed emergency plans are practicable;
    (B) These agencies are committed to participating in any further 
development of the plans, including any required field demonstrations; 
and
    (C) These agencies are committed to executing their 
responsibilities under the plans in the event of an emergency.
    (ii) If certifications cannot be obtained after sustained, good 
faith efforts by the applicant, then the application must contain 
information, including a utility plan, sufficient to show that the 
proposed plans provide reasonable assurance that adequate protective 
measures can and will be taken in the event of a radiological emergency 
at the site.
    (10) Organization. A description of the applicant's organizational 
structure, allocations of responsibilities and authorities, and 
personnel qualifications requirements for operation.
    (11) Maintenance program. A description of a maintenance program 
under Sec.  53.715.
    (12) Quality assurance. A description of the QAP under Sec.  
53.865.
    (13) Radiation protection program. A radiation protection program 
description under Sec.  53.850.
    (14) Security program. A physical security plan that describes how 
the applicant will comply with Sec.  53.860 (and 10 CFR part 11, if 
applicable, including the identification and description of jobs as 
required by Sec.  11.11(a) of this chapter, at the proposed facility). 
The plan must list tests, inspections, audits, and other means to be 
used to demonstrate compliance with the requirements of 10 CFR parts 11 
and 73, if applicable.
    (15) Safeguards contingency plan. A safeguards contingency plan in 
accordance with the criteria set forth in appendix C to 10 CFR part 73. 
The safeguards contingency plan must include plans for dealing with 
threats, thefts, and radiological sabotage, as defined in 10 CFR part 
73, relating to the SNM and nuclear facilities licensed under this 
chapter and in the applicant's possession and control. Each

[[Page 87098]]

application for this type of license must include the information 
contained in the applicant's safeguards contingency plan.\1\ 
(Implementing procedures required for this plan need not be submitted 
for approval.)
    (16) Security training and qualification. A training and 
qualification plan that describes how the applicant will demonstrate 
compliance with the criteria set forth in Sec.  73.100 of this chapter 
or appendix B to 10 CFR part 73.
    (17) Cybersecurity plan. A cybersecurity plan in accordance with 
the criteria set forth in Sec.  73.54 or Sec.  73.110 of this chapter.
    (18) Security, safeguards and cybersecurity plan implementation. A 
description of the implementation of the physical security plan, 
safeguards contingency plan, training and qualification plan, and 
cybersecurity plan. Each applicant who prepares a physical security 
plan, a safeguards contingency plan, a training and qualification plan, 
or a cybersecurity plan must protect the plans and other related 
Safeguards Information against unauthorized disclosure in accordance 
with the requirements of Sec. Sec.  73.21 and 73.22 of this chapter.
    (19) Fire protection program. A description of the fire protection 
program under Sec.  53.875.
    (20) Inservice inspection/inservice testing program. Descriptions 
of inservice inspection and inservice testing programs under Sec.  
53.880.
    (21) [Reserved]
    (22) [Reserved]
    (23) General employee training. A description of the training 
program required to demonstrate compliance with Sec.  53.830 and its 
implementation.
    (24) Fitness-for-duty program. A description of the FFD program 
under part 26 of this chapter and its implementation.
    (25) Technical specifications. Proposed technical specifications 
prepared in accordance with the requirements of Sec.  53.710(a).
    (b) If there are SSCs of the plant for which research and 
development is necessary to confirm the adequacy of their design, a 
report which documents the resolution of any safety questions 
associated with such SSCs.
    (c) A description of how the performance of each safety design 
feature has been demonstrated capable of fulfilling functional design 
criteria considering interdependent effects through either analysis, 
appropriate test programs, prototype testing, operating experience, or 
a combination thereof, in accordance with Sec.  53.440(a).
    (d) If the COL application references an early site permit, then 
the following requirements apply:
    (1) The FSAR need not contain information or analyses submitted to 
the Commission in connection with the early site permit provided that 
the FSAR must either include or incorporate by reference the early site 
permit Site Safety Analysis Report and contain, in addition to the 
information and analyses otherwise required, information sufficient to 
demonstrate that the design of the facility falls within the site 
characteristics and design parameters specified in the early site 
permit.
    (2) If the FSAR does not demonstrate that design of the facility 
falls within the site characteristics and design parameters, the 
application must include a request for a variance that complies with 
the requirements of Sec. Sec.  53.1188(d) and 53.1437.
    (3) The FSAR must demonstrate that all terms and conditions that 
have been included in the early site permit will be satisfied by the 
date of issuance of the COL. Any terms or conditions of the early site 
permit that could not be met by the time of issuance of the COL must be 
set forth as terms or conditions of the COL.
    (4) If the early site permit approves complete and integrated 
emergency plans, or major features of emergency plans, then the FSAR 
must include any new or additional information that updates and 
corrects the information that was provided under Sec.  53.1146(b)(2) 
and discuss whether the new or additional information materially 
changes the bases for compliance with the applicable requirements. The 
application must identify changes to the emergency plans or major 
features of emergency plans that have been incorporated into the 
proposed facility emergency plans and that constitute or would 
constitute a change in an emergency plan that results in reducing the 
licensee's capability to perform an emergency planning function in the 
event of a radiological emergency.
    (5) If complete and integrated emergency plans are approved as part 
of the early site permit, new certifications meeting the requirements 
of paragraph (a)(9)(i) of this section are not required.
    (e) If the COL application references a standard design approval, 
then the following requirements apply:
    (1) The FSAR need not contain information or analyses submitted to 
the Commission in connection with the design approval, provided, 
however, that the FSAR must either include or incorporate by reference 
the standard design approval FSAR and must contain, in addition to the 
information and analyses otherwise required, information sufficient to 
demonstrate that the characteristics of the site fall within the site 
parameters specified in the design approval. In addition, the plant-
specific PRA information must use the PRA information for the design 
approval and must be updated to account for site specific design 
information and any design changes or departures.
    (2) The FSAR must demonstrate that all terms and conditions that 
have been included in the design approval will be satisfied by the date 
of issuance of the COL.
    (f) If the COL application references a standard design 
certification, then the following requirements apply:
    (1) The FSAR need not contain information or analyses submitted to 
the Commission in connection with the standard design certification, 
provided, however, that the FSAR must either include or incorporate by 
reference the standard design certification FSAR and must contain, in 
addition to the information and analyses otherwise required, 
information sufficient to demonstrate that the site characteristics 
fall within the site parameters specified in the standard design 
certification. In addition, the plant-specific PRA information must use 
the PRA information for the standard design certification and must be 
updated to account for site-specific design information and any design 
changes or departures.
    (2) The FSAR must demonstrate that the interface requirements 
established for the design under Sec.  53.1239(a)(24) have been met.
    (3) The FSAR must demonstrate that all requirements and 
restrictions set forth in the referenced standard design certification 
rule must be satisfied by the date of issuance of the COL. Any 
requirements and restrictions set forth in the referenced standard 
design certification rule that could not be satisfied by the time of 
issuance of the COL, must be set forth as terms or conditions of the 
COL.
    (g) If the COL application references the use of one or more 
manufactured reactors licensed under Sec.  53.1270, then the following 
requirements apply:
    (1) The FSAR need not contain information or analyses submitted to 
the Commission in connection with the ML, provided, however, that the 
FSAR must either include or incorporate by reference the ML FSAR and 
must contain, in addition to the information and analyses otherwise 
required, information sufficient to demonstrate that the site 
characteristics fall within the site parameters specified in the ML.

[[Page 87099]]

In addition, the plant-specific PRA information must use the PRA 
information for the manufactured reactor and must be updated to account 
for site-specific design information and any design changes or 
departures.
    (2) The FSAR must demonstrate that the interface requirements 
established for the design have been met.
    (3) The FSAR must demonstrate that all terms and conditions that 
have been included in the ML will be satisfied by the date of issuance 
of the COL. Any terms or conditions of the ML that could not be met by 
the time of issuance of the COL, must be set forth as terms or 
conditions of the COL.
    (h) Each applicant for a COL under this part must protect 
Safeguards Information against unauthorized disclosure in accordance 
with the requirements in Sec. Sec.  73.21 and 73.22 of this chapter, as 
applicable.

    \1\ A physical security plan that contains all the information 
required in both Sec.  73.55 or Sec.  73.100 of this chapter and 
appendix C to 10 CFR part 73 demonstrates compliance with the 
requirement for a contingency plan.


Sec.  53.1419  Contents of applications for combined licenses; other 
application content.

    (a) In addition to the FSAR, the application must also include the 
following:
    (1) Environmental report. (i) An environmental report either in 
accordance with Sec.  51.50(c) of this chapter if an LWA under Sec.  
53.1130 is not requested in conjunction with the COL application, or in 
accordance with Sec. Sec.  51.49 and 51.50(c) of this chapter if an LWA 
is requested in conjunction with the COL application; or
    (ii) If the applicant wishes to request that an LWA under Sec.  
53.1130 be issued before issuance of the COL, the information otherwise 
required by Sec.  53.1130, in accordance with either Sec.  2.101(a)(1) 
through (a)(4), or Sec.  2.101(a)(9) of this chapter;
    (2) Availability controls (if not included in the FSAR). A 
description of the controls on plant operations, including availability 
controls, to provide reasonable confidence of safe operation and that 
the configurations and special treatments for SR SSCs and NSRSS SSCs 
provide the capabilities and reliabilities required to satisfy the 
safety criteria of Sec.  53.220, or more restrictive alternative 
criteria adopted under Sec.  53.470, if not addressed by Technical 
Specifications under Sec.  53.1416(a)(25); and
    (3) Inspections, tests, analyses, and acceptance criteria. The 
proposed inspections, tests, and analyses, including those applicable 
to emergency planning, that the licensee must perform, and the 
acceptance criteria that are necessary and sufficient to provide 
reasonable assurance that, if the inspections, tests, and analyses are 
performed and the acceptance criteria met, the facility has been 
constructed and will be operated in conformity with the COL, the 
provisions of the Act, and the Commission's rules and regulations.
    (i) If the application references an early site permit with ITAAC, 
the early site permit ITAAC must apply to those aspects of the COL 
which are approved in the early site permit.
    (ii) If the application references a standard design certification, 
the ITAAC contained in the certified design must apply to those 
portions of the facility design which are approved in the standard 
design certification.
    (iii) If the application references an ML, the ITAAC contained in 
the ML must apply to those portions of the facility design which are 
approved in the ML.
    (iv) If the application references an early site permit with ITAAC, 
a standard design certification, an ML, or combination thereof, the 
application may include a notification that a required inspection, 
test, or analysis in the ITAAC has been successfully completed and that 
the corresponding acceptance criterion has been met. The Federal 
Register notification required by Sec.  53.1422 of this chapter must 
indicate that the application includes this notification.
    (b) [Reserved]


Sec.  53.1422  Review of applications.

    (a) Standards for review of applications. Applications filed under 
this part will be reviewed according to the standards set out in this 
part and 10 CFR parts 20, 51, 73, and 140.
    (b) Administrative review of applications; hearings. A proceeding 
on a COL is subject to all applicable procedural requirements contained 
in 10 CFR part 2, including the requirements for docketing (Sec.  2.101 
of this chapter) and issuance of a notice of hearing (Sec.  2.104 of 
this chapter). If an applicant requests a Commission finding on certain 
ITAAC with the issuance of the COL, then those ITAAC will be identified 
in the notice of hearing. All hearings on COLs are governed by the 
procedures contained in 10 CFR part 2.


Sec.  53.1425  Finality of referenced NRC approvals.

    If the application for a COL under this part references an early 
site permit, standard design certification rule, standard design 
approval, or ML, issued under this part, the scope and nature of 
matters resolved for the application and any COL issued are governed by 
the relevant provisions addressing finality, including Sec. Sec.  
53.1188, 53.1221, 53.1263, and 53.1288.


Sec.  53.1431  Referral to the Advisory Committee on Reactor 
Safeguards.

    The Commission must refer a copy of the application to the ACRS. 
The ACRS must report on those portions of the application that concern 
safety and must apply the standards referenced in Sec.  53.1422, in 
accordance with the finality provisions in Sec.  53.1425.


Sec.  53.1434  Authorization to conduct limited work authorization 
activities.

    (a) If the application for a COL under this part does not reference 
an early site permit which authorizes the holder to perform the 
activities under Sec.  53.1130(b), the applicant may not perform those 
activities without obtaining the separate authorization required by 
Sec.  53.1130(a). Authorization may be granted only after the presiding 
officer in the proceeding on the application has made the findings and 
determination required by Sec.  53.1130(b)(1)(ii) and (b)(1)(iv), and 
the Director, Office of Nuclear Reactor Regulation makes the 
determination required by Sec.  53.1130(b)(1)(iii).
    (b) If, after an applicant has performed the activities permitted 
by a LWA issued under Sec.  53.1130, the application for the COL is 
withdrawn or denied, then the applicant must implement the approved 
site redress plan.


Sec.  53.1437  Exemptions, departures, and variances.

    (a) An applicant for a COL, or any amendment to a COL, may include 
in the application a request for an exemption from one or more of the 
Commission's regulations.
    (1) If the request is for an exemption from any part of a 
referenced standard design certification rule, the Commission may grant 
the request if it determines that the exemption complies with any 
exemption provisions of the referenced standard design certification 
rule, or with Sec.  53.1263 if there are no applicable exemption 
provisions in the referenced standard design certification rule.
    (2) For all other requests for exemptions, the Commission may grant 
a request if it determines that the exemption complies with Sec.  
53.080.
    (b) An applicant for a COL who has filed an application referencing 
an early site permit issued under Sec.  53.1158 may include in the 
application a request for a variance from one or more site 
characteristics, design parameters, or

[[Page 87100]]

terms and conditions of the permit, or from the Site Safety Analysis 
Report. In determining whether to grant the variance, the Commission 
must apply the same technically relevant criteria as were applicable to 
the application for the original or renewed site permit. Once a COL 
referencing an early site permit is issued, variances from the early 
site permit will not be granted for that CP or COL.
    (c) An applicant for a COL who has filed an application referencing 
use of a manufactured reactor may include in the application a request 
for a departure from one or more design characteristics, site 
parameters, terms and conditions, or approved design of the 
manufactured reactor under the ML issued under Sec.  53.1287. The 
Commission may grant such a request only if it determines that the 
departure will comply with the requirements of Sec.  53.080, and that 
the special circumstances outweigh any decrease in safety that may 
result from the reduction in standardization caused by the departure.
    (d) Issuance of a variance under paragraph (b) of this section or a 
departure under paragraph (c) of this section is subject to litigation 
during the COL proceeding in the same manner as other issues material 
to that proceeding.


Sec.  53.1440  Issuance of combined licenses.

    (a)(1) After conducting a hearing under Sec.  53.1422(b) and 
receiving the report submitted by the ACRS, the Commission may issue a 
COL if the Commission finds that--
    (i) The applicable standards and requirements of the Act and the 
Commission's regulations have been met;
    (ii) Any required notifications to other agencies or bodies have 
been duly made;
    (iii) There is reasonable assurance that the facility will be 
constructed and will operate in conformity with the license, the 
provisions of the Act, and the Commission's regulations;
    (iv) The applicant is technically and financially qualified to 
engage in the activities authorized; however, no finding of financial 
qualification is necessary for an electric utility applicant for a COL;
    (v) Issuance of the license will not be inimical to the common 
defense and security or to the health and safety of the public; and
    (vi) The findings required by subpart A of 10 CFR part 51 have been 
made.
    (2) The Commission may also find, at the time it issues the COL, 
that certain acceptance criteria in one or more of the ITAAC in a 
referenced early site permit, standard design certification, or ML have 
been met. This finding will finally resolve that those acceptance 
criteria have been met, those acceptance criteria will be deemed to be 
excluded from the COL, and findings under Sec.  53.1452(g) with respect 
to those acceptance criteria are unnecessary.
    (b) The Commission must identify within the COL the inspections, 
tests, and analyses, including those applicable to emergency planning, 
that the licensee must perform, and the acceptance criteria that, if 
met, are necessary and sufficient to provide reasonable assurance that 
the facility has been constructed and will be operated in conformity 
with the license, the provisions of the Act, and the Commission's rules 
and regulations.
    (c) A COL must contain the terms and conditions, including 
technical specifications, as the Commission deems necessary and 
appropriate.


Sec.  53.1443  Finality of combined licenses.

    (a) After issuance of a COL, the Commission may not modify, add, or 
delete any term or condition of the COL, the design of the facility, 
the ITAAC contained in the license that are not derived from a 
referenced standard design certification or ML, except under the 
provisions of Sec.  53.1452 or Sec.  53.1590.
    (b) If the COL does not reference a standard design certification 
or use of a manufactured reactor under an ML issued under Sec.  
53.1287, then a licensee may make changes in the facility as described 
in the FSAR (as updated) and make changes in the procedures as 
described in the FSAR (as updated) under the applicable change 
processes in Sec.  53.1550.
    (c) If the COL references a certified design, then--
    (1) Changes to or departures from information within the scope of 
the referenced standard design certification rule are subject to the 
applicable change processes in that rule; and
    (2) Changes that are not within the scope of the referenced 
standard design certification rule are subject to the applicable change 
processes in subpart I of this part, unless they also involve changes 
to or noncompliance with information within the scope of the referenced 
standard design certification rule. In these cases, the applicable 
provisions of this section and the standard design certification rule 
apply.
    (d) If the COL references use of a manufactured reactor under an ML 
issued under this part, then--
    (1) Changes to or departures from information within the scope of 
the manufactured reactor's design are subject to the change processes 
in Sec.  53.1288; and
    (2) Changes that are not within the scope of the manufactured 
reactor's design are subject to the applicable change processes in 
subpart I.
    (e) The Commission may issue and make immediately effective any 
amendment to a COL upon a determination by the Commission that the 
amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person. The amendment may be issued and made 
immediately effective in advance of the holding and completion of any 
required hearing. The amendment will be processed under the procedures 
specified in Sec.  53.1515.
    (f) Any modification to, addition to, or deletion from the terms 
and conditions of a COL, including any modification to, addition to, or 
deletion from the inspections, tests, and analyses, or related 
acceptance criteria contained in the license is a proposed amendment to 
the license. There must be an opportunity for a hearing on the 
amendment.


Sec.  53.1449  Inspection during construction.

    (a) Licensee schedule for inspections, tests, or analyses. The 
licensee must submit to the NRC, no later than 1 year after issuance of 
the COL or at the start of construction as defined at Sec.  53.020, 
whichever is later, its schedule for completing the inspections, tests, 
or analyses in the ITAAC. The licensee must submit updates to the ITAAC 
schedules every 6 months thereafter and, within 1 year of its scheduled 
date for initial loading of fuel (or, for a fueled manufactured 
reactor, within 1 year of its scheduled date for initiating the 
physical removal of any one of the independent physical mechanisms to 
prevent criticality required under Sec.  53.620(d)(1)), the licensee 
must submit updates to the ITAAC schedule every 30 days until the final 
notification is provided to the NRC under paragraph (c)(1) of this 
section.
    (b) Licensee and applicant conduct of activities subject to ITAAC. 
With respect to activities subject to an ITAAC, an applicant for a COL 
may proceed at its own risk with design and procurement activities, and 
a licensee may proceed at its own risk with design, procurement, 
construction, and preoperational activities, even though the NRC may 
not have found that any one of the prescribed acceptance criteria are 
met.
    (c) Licensee notifications. (1) ITAAC closure notification. The 
licensee must notify the NRC that prescribed inspections, tests, and 
analyses have been performed and that the prescribed acceptance 
criteria are met. The

[[Page 87101]]

notification must contain sufficient information to demonstrate that 
the prescribed inspections, test, and analyses have been performed and 
that the prescribed acceptance criteria are met.
    (2) ITAAC post-closure notifications. Following the licensee's 
ITAAC closure notifications under paragraph (c)(1) of this section 
until the Commission makes the finding under Sec.  53.1452(g), the 
licensee must notify the NRC, in a timely manner, of new information 
that materially alters the basis for determining that either 
inspections, tests, or analyses were performed as required, or that 
acceptance criteria are met. The notification must contain sufficient 
information to demonstrate that, notwithstanding the new information, 
the prescribed inspections, tests, and analyses have been performed as 
required, and the prescribed acceptance criteria are met.
    (3) Uncompleted ITAAC notification. If the licensee has not 
provided, by the date 225 days before the scheduled date for initial 
loading of fuel (or, for a fueled manufactured reactor, by the date 225 
days before the scheduled date for initiating the physical removal of 
any one of the independent physical mechanisms to prevent criticality 
required under Sec.  53.620(d)(1)), the notification required by 
paragraph (c)(1) of this section for all ITAAC, then the licensee must 
notify the NRC that the prescribed inspections, tests, or analyses for 
all uncompleted ITAAC will be performed and that the prescribed 
acceptance criteria will be met prior to operation. The notification 
must be provided no later than the date 225 days before the scheduled 
date for initial loading of fuel (or, for a fueled manufactured 
reactor, no later than the date 225 days before the scheduled date for 
initiating the physical removal of any one of the independent physical 
mechanisms to prevent criticality required under Sec.  53.620(d)(1)), 
and must provide sufficient information to demonstrate that the 
prescribed inspections, tests, or analyses will be performed and the 
prescribed acceptance criteria for the uncompleted ITAAC will be met, 
including, but not limited to, a description of the specific procedures 
and analytical methods to be used for performing the prescribed 
inspections, tests, and analyses and determining that the prescribed 
acceptance criteria are met.
    (4) All ITAAC complete notification. The licensee must notify the 
NRC that all ITAAC are complete.
    (d) Licensee determination of noncompliance with ITAAC. (1) In the 
event that an activity is subject to an ITAAC derived from a referenced 
standard design certification and the licensee has not demonstrated 
that the prescribed acceptance criteria are met, the licensee may take 
corrective actions to successfully complete that ITAAC or request an 
exemption from the standard design certification ITAAC, as applicable. 
A request for an exemption must also be accompanied by a request for a 
license amendment under subpart I.
    (2) In the event that an activity is subject to an ITAAC not 
derived from a referenced standard design certification and the 
licensee has not demonstrated that the prescribed acceptance criteria 
are met, the licensee may take corrective actions to successfully 
complete that ITAAC or request a license amendment under subpart I.
    (e) NRC inspection, publication of notices, and availability of 
licensee notifications. The NRC must ensure that the prescribed 
inspections, tests, and analyses in the ITAAC are performed.
    (1) At appropriate intervals until the last date for submission of 
requests for hearing under Sec.  53.1452, the NRC must publish notices 
in the Federal Register of the NRC staff's determination of the 
successful completion of inspections, tests, and analyses.
    (2) The NRC must make publicly available the licensee notifications 
under paragraph (c) of this section. The NRC must, no later than the 
date of publication of the notice of intended operation required by 
Sec.  53.1452(a), make publicly available those licensee notifications 
under paragraph (c) of this section that have been submitted to the NRC 
at least 7 days before that notice.


Sec.  53.1452  Operation under a combined license.

    (a) The licensee must notify the NRC of its scheduled date for 
initial loading of fuel no later than 270 days before the scheduled 
date and must notify the NRC of updates to its schedule every 30 days 
thereafter.\1\ Not less than 180 days before the date scheduled for 
initial loading of fuel into a plant by a licensee that has been issued 
a COL under this part, the Commission must publish notice of intended 
operation in the Federal Register.\2\ The notice must provide that any 
person whose interest may be affected by operation of the plant may, 
within 60 days, request that the Commission hold a hearing on whether 
the facility as constructed complies, or on completion will comply, 
with the acceptance criteria in the COL, except that a hearing must not 
be granted for those ITAAC that the Commission found were met under 
Sec.  53.1440(a)(2).
    (b) A request for hearing under paragraph (a) of this section must 
show, prima facie--
    (1) That one or more of the acceptance criteria of the ITAAC in the 
COL have not been, or will not be, met; and
    (2) The specific operational consequences of nonconformance that 
would be contrary to providing reasonable assurance of adequate 
protection of the public health and safety.
    (c) The Commission, acting as the presiding officer, must determine 
whether to grant or deny the request for hearing under the applicable 
requirements of Sec.  2.309 of this chapter. If the Commission grants 
the request, the Commission, acting as the presiding officer, must 
determine whether during a period of interim operation there will be 
reasonable assurance of adequate protection to the public health and 
safety. The Commission's determination must consider the petitioner's 
prima facie showing and any answers thereto. If the Commission 
determines there is such reasonable assurance, it must allow operation 
during an interim period under the COL.
    (d) The Commission, in its discretion, must determine appropriate 
hearing procedures, whether informal or formal adjudicatory, for any 
hearing under paragraph (a) of this section, and must state its reasons 
therefore.
    (e) The Commission must, to the maximum possible extent, render a 
decision on issues raised by the hearing request within 180 days of the 
publication of the notice provided by paragraph (a) of this section or 
by the anticipated date for initial loading of fuel into the reactor, 
whichever is later.
    (f) A petition to modify the terms and conditions of the COL will 
be processed as a request for action under Sec.  2.206 of this chapter. 
The petitioner must file the petition with the Secretary of the 
Commission. Before the licensed activity allegedly affected by the 
petition (fuel loading, low power testing, etc.) commences, the 
Commission must determine whether any immediate action is required. If 
the petition is granted, then an appropriate order will be issued. Fuel 
loading and operation under the COL will not be affected by the 
granting of the petition unless the order is made immediately 
effective.
    (g) The licensee must not operate the facility until the Commission 
makes a finding that the acceptance criteria in the COL are met, except 
for those acceptance criteria that the Commission found were met under 
Sec.  53.1440(a)(2). If the COL is for a modular design, each

[[Page 87102]]

reactor unit may require a separate finding as construction proceeds.
    (h) After the Commission has made the finding in paragraph (g) of 
this section, the ITAAC do not, by virtue of their inclusion in the 
COL, constitute regulatory requirements either for licensees or for 
renewal of the license; except for the specific ITAAC for which the 
Commission has granted a hearing under paragraph (a) of this section, 
all ITAAC expire upon final Commission action in the proceeding. 
However, subsequent changes to the facility or procedures described in 
the FSAR (as updated) must comply with the requirements in Sec.  
53.1443(e) or (f), as applicable.

    \1\ For licensees installing fueled manufactured reactors under 
a COL, the COL holder must instead notify the NRC of its scheduled 
date for initiating the physical removal of any one of the 
independent physical mechanisms to prevent criticality required 
under Sec.  53.620(d)(1) no later than 270 days before the scheduled 
date and must notify the NRC of updates to its schedule every 30 
days thereafter.
    \2\ For licensees installing fueled manufactured reactors under 
a COL, the Commission must instead publish notice of intended 
operation in the Federal Register not less than 180 days before the 
date scheduled for initiating the physical removal of any one of the 
independent physical mechanisms to prevent criticality required 
under Sec.  53.620(d)(1).


Sec.  53.1455  Duration of combined license.

    A COL is issued for a specified period not to exceed 40 years from 
the date on which the Commission makes a finding that acceptance 
criteria are met under Sec.  53.1452(g) or allowing operation during an 
interim period under the COL under Sec.  53.1452(c).


Sec.  53.1456  Transfer of a combined license.

    A COL may be transferred under Sec.  53.1570.


Sec.  53.1458  Application for renewal.

    The filing of an application for a renewed license must be in 
accordance with Sec.  53.1595.


Sec.  53.1461  Continuation of combined license.

    Each COL for a facility that has permanently ceased operations 
continues in effect beyond the expiration date to authorize ownership 
and possession of the facility until the Commission notifies the 
licensee in writing that the license is terminated. During this period 
of continued effectiveness, the licensee must--
    (a) Take actions necessary to decommission and decontaminate the 
facility and continue to maintain the facility, including, where 
applicable, the storage, control and maintenance of the spent fuel, in 
a safe condition; and
    (b) Conduct activities in accordance with all other restrictions 
applicable to the facility in accordance with the NRC's regulations and 
the provisions of the COL for the facility.


Sec.  53.1470  Standardization of commercial nuclear plant designs: 
licenses to construct and operate nuclear power reactors of identical 
design at multiple sites.

    (a) Except as otherwise specified in this section, the provisions 
of this section apply to CP, OL, and COL applications for commercial 
nuclear plants of identical design (the ``common design'') under this 
part.
    (b) Each application for a CP, OL, or COL submitted pursuant to 
this section must be submitted as specified in Sec. Sec.  53.1300, 
53.1360, or 53.1410, respectively, and Sec.  2.101 of this chapter. 
Each application must state that the applicant wishes to construct a 
facility identical to a facility proposed for one or more sites other 
than the applicant's (the ``common design''), and the applicant wishes 
to have the application considered under this section. Each application 
must list each of the other applications to be treated together under 
this section.
    (c) Each application must include the information required by the 
applicable sections of this subpart, provided however, that the 
application must identify the common design, and, if applicable, 
reference a standard design certification or standard design approval 
under this part, or the use of a reactor manufactured under this part. 
The FSAR for each application must either incorporate by reference or 
include the final safety analysis of the common design, including, if 
applicable, the FSAR for the referenced standard design certification, 
standard design approval, or the manufactured reactor.
    (d) Each application submitted pursuant to this section must 
contain an environmental report under Sec. Sec.  53.1312(a)(1), 
53.1372(a), or 53.1419(a)(1), as applicable, that complies with the 
applicable provisions of 10 CFR part 51, provided, however, that the 
application may incorporate by reference a single environmental report 
on the environmental impacts of the common design that are applicable 
to each site.
    (e) Upon a determination that each application is acceptable for 
docketing under Sec.  2.101 of this chapter, each application will be 
docketed and a notice of docketing for each application will be 
published in the Federal Register, under Sec.  2.104 of this chapter, 
provided, however, that the notice must state that the application will 
be processed under the provisions of this section and subpart D of 10 
CFR part 2. At the discretion of the Commission, a single notice of 
docketing for multiple applications may be published in the Federal 
Register.
    (f) The NRC must prepare an environmental assessment or draft and 
final environmental impact statements for each of the applications 
under 10 CFR part 51. Scoping under Sec. Sec.  51.28 and 51.29 of this 
chapter for each of the license applications may be conducted 
simultaneously and joint scoping may be conducted with respect to the 
environmental issues relevant to the common design. If the applications 
reference a standard design certification, then the environmental 
assessment or environmental impact statement for each of the 
applications must incorporate by reference the standard design 
certification environmental assessment. If the applications do not 
reference a standard design certification, then the NRC must prepare 
environmental assessments or draft and final supplemental environmental 
impact statements which address severe accident mitigation design 
alternatives for the common design, which must be incorporated by 
reference into the environmental assessment or environmental impact 
statement prepared for each application. Scoping under Sec. Sec.  51.28 
and 51.29 of this chapter for the supplemental environmental impact 
statement may be conducted simultaneously and may be part of the 
scoping for each of the applications.
    (g) The ACRS must report on each of the applications as required by 
the applicable sections of this subpart. Each report must be limited to 
those safety matters for each application that are not relevant to the 
common design. In addition, the ACRS must separately report on the 
safety of the common design, provided, however, that the report need 
not address the safety of a referenced standard design certification or 
reactor manufactured under this part.
    (h) The Commission must designate a presiding officer to conduct 
the proceeding with respect to the health and safety, common defense 
and security, and environmental matters relating to the common design 
and affecting at least two applications. The hearing will be governed 
by the applicable provisions of subparts A, C, G, L, N, and O of 10 CFR 
part 2 relating to applications for CPs, OLs, and COLs. The presiding 
officer must issue a partial initial decision on the common design.
    (i) If the design for the power reactor(s) proposed in a particular

[[Page 87103]]

application is not identical to the others, that application may not be 
processed under this section and subpart D of 10 CFR part 2.
    (j) As used in this section, the design of a nuclear power reactor 
included in a single referenced Safety Analysis Report means the design 
of those SSCs important to radiological health and safety and the 
common defense and security.

Subpart I--Maintaining and Revising Licensing-Basis Information


Sec.  53.1500  Licensing-basis information.

    This subpart provides the requirements for each holder of a license 
for a commercial nuclear plant licensed under this part to maintain 
licensing-basis information as defined in Sec.  53.020; evaluate 
changes to site characteristics, plant design features, and 
programmatic controls to determine needed approvals and revisions; and 
submit appropriate updates to the U.S. Nuclear Regulatory Commission 
(NRC).


Sec.  53.1502  Specific terms and conditions of licenses.

    (a) Each license issued under this part is subject to the 
provisions of the Atomic Energy Act of 1954, as amended, (the Act) and 
to all rules, regulations, and orders of the Commission. The terms and 
conditions of the license will be subject to amendment, revision, or 
modification, by reason of amendments of the Act or by reason of rules, 
regulations, and orders issued in accordance with the terms of the Act.
    (b) Each license issued under this part must be subject to all 
conditions imposed as a matter of law by sections 401(a)(2) and 401(d) 
of the Federal Water Pollution Control Act, as amended (33 U.S.C.A. 
1341(a)(2) and (d)).
    (c) A holder of an operating license (OL) or combined license (COL) 
under this part may take reasonable action that departs from a license 
condition or a technical specification included in a license issued 
under this part in a national security emergency established by a law 
enacted by the Congress or by an order or directive issued by the 
President pursuant to statutes or the Constitution of the United 
States. The authority under this paragraph must be exercised in 
accordance with law, including section 57e of the Act, and is in 
addition to the authority granted under Sec.  53.740(h), which remains 
in effect unless otherwise directed by the Commission during a national 
security emergency. The authority under this paragraph may be 
exercised--
    (1) When this action is immediately needed to implement national 
security objectives as designated by the national command authority 
through the Commission; and
    (2) No action consistent with license conditions and technical 
specifications that can satisfy national security objectives is 
immediately apparent.
    (d)(1) If the NRC finds that the state of emergency preparedness 
does not provide reasonable assurance that adequate protective measures 
can and will be taken in the event of a radiological emergency 
(including findings based on requirements of 10 CFR part 50, appendix 
E, section IV.D.3) and if the deficiencies (including deficiencies 
based on requirements of 10 CFR part 50, appendix E, section IV.D.3) 
are not corrected within 4 months of that finding, the Commission will 
determine whether the facility must be shut down or cease operations 
until such deficiencies are remedied or whether other enforcement 
action is appropriate. In determining whether a shutdown or other 
enforcement action is appropriate, the Commission will take into 
account, among other factors, whether the licensee can demonstrate to 
the Commission's satisfaction that the deficiencies in the plan are not 
significant for the plant in question, or that adequate interim 
compensating actions have been or will be taken promptly, or that that 
there are other compelling reasons for continued operation.
    (2) If the planning standards for radiological emergency 
preparedness apply to offsite emergency response plans, or if the 
planning activities in Sec.  50.160(b)(1)(iv)(B) apply, then the NRC 
will base its finding on a review of the Federal Emergency Management 
Agency findings and determinations as to whether State, participating 
Tribal and local emergency plans are adequate and capable of being 
implemented, and on the NRC assessment as to whether the licensee's 
emergency plans are adequate and capable of being implemented. Nothing 
in this paragraph must be construed as limiting the authority of the 
Commission to take action under any other regulation or authority of 
the Commission or at any time other than that specified in this 
paragraph.


Sec.  53.1505  Changes to licensing-basis information requiring prior 
NRC approval.

    (a) Sections 53.1510 through 53.1520 provide the process for a 
licensee to request and the NRC to issue amendments to licenses, 
including any conditions contained therein, technical specifications or 
other attachments to a license, and any orders issued by the NRC 
modifying a license. Sections 53.1525 and 53.1530 govern proposed 
changes to a commercial nuclear plant referencing a certified design or 
manufacturing license (ML).
    (b) A licensee may propose changing licensing-basis information 
established by NRC regulations by requesting an exemption in accordance 
with Sec.  53.080.


Sec.  53.1510  Application for amendment of license.

    Whenever a holder of a license under this part desires to amend the 
license, an application for an amendment must be filed with the 
Commission, as specified in Sec.  53.040, that fully describes the 
changes desired and, following as far as applicable, the form 
prescribed for original applications. Applications for amendments 
involving changes to plant structures, systems, and components (SSCs), 
programmatic controls, or the role of plant personnel must include an 
assessment of the changes in relation to the safety requirements in 
subpart B of this part and the analyses requirements of Sec.  53.450 as 
applicable, an analysis of whether the amendment involves no 
significant hazards consideration using the standards in Sec.  53.1520, 
and a consideration of environmental factors.


Sec.  53.1515  Public notices; State consultation.

    The Commission will use the following procedures for an application 
requesting an amendment to an OL or COL issued under this part.
    (a) Public notices. (1)(i) The Commission may publish in the 
Federal Register under Sec.  2.105 of this chapter an individual notice 
of proposed action for an amendment for which it makes a proposed 
determination that no significant hazards consideration is involved, 
or, at least once every 30 days, publish a periodic Federal Register 
notice of proposed actions, which identifies each amendment issued and 
each amendment proposed to be issued since the last such periodic 
notice, or it may publish both such notices.
    (ii) For each amendment proposed to be issued, the notice will
    (A) Contain the staff's proposed determination under the standards 
in Sec.  53.1520;
    (B) Provide a brief description of the amendment and of the 
facility involved;
    (C) Solicit public comments on the proposed determination; and
    (D) Provide for a 30-day comment period.
    (iii) The comment period will begin on the day after the date of 
the publication of the first notice, and, normally, the amendment will 
not be granted until after this comment period expires.

[[Page 87104]]

    (2) The Commission may inform the public about the final 
disposition of an amendment request for which it has made a proposed 
determination of no significant hazards consideration either by issuing 
an individual notice of issuance under Sec.  2.106 of this chapter or 
by publishing such a notice in its periodic system of Federal Register 
notices. In either event, it will not make and will not publish a final 
determination of no significant hazards consideration unless it 
receives a request for a hearing on that amendment request.
    (3) Where the Commission makes a final determination that no 
significant hazards consideration is involved and that the amendment 
should be issued, the amendment will be effective on issuance, even if 
adverse public comments have been received and even if an interested 
person meeting the provisions for intervention called for in Sec.  
2.309 of this chapter has filed a request for a hearing. The Commission 
need hold any required hearing only after it issues an amendment, 
unless it determines that a significant hazards consideration is 
involved, in which case the Commission will provide an opportunity for 
a prior hearing.
    (4) Where the Commission finds that an emergency situation exists, 
in that failure to act in a timely way would result in derating or 
shutdown of a commercial nuclear reactor, or in prevention of either 
resumption of operation or of increase in power output up to the 
plant's licensed power level, it may issue a license amendment 
involving no significant hazards consideration without prior notice and 
opportunity for a hearing or for public comment. In such a situation, 
the Commission will not publish a notice of proposed determination on 
no significant hazards consideration but will publish a notice of 
issuance under Sec.  2.106 of this chapter providing for opportunity 
for a hearing and for public comment after issuance. The Commission 
expects its licensees to apply for license amendments in a timely 
fashion. It will decline to dispense with notice and comment on the 
determination of no significant hazards consideration if it determines 
that the licensee has abused the emergency provision by failing to make 
timely application for the amendment and thus itself creating the 
emergency. Whenever an emergency situation exists, a licensee 
requesting an amendment must explain why this emergency situation 
occurred and why it could not avoid this situation, and the Commission 
will assess the licensee's reasons for failing to file an application 
sufficiently in advance of that event.
    (5) Where the Commission finds that exigent circumstances exist, in 
that a licensee and the Commission must act quickly and that time does 
not permit the Commission to publish a Federal Register notice allowing 
30 days for prior public comment, and it also determines that the 
amendment involves no significant hazards considerations, it--
    (i)(A) Will either issue a Federal Register notice providing notice 
of an opportunity for hearing and allowing at least 2 weeks from the 
date of the notice for prior public comment; or
    (B) Will use local media to provide reasonable notice to the public 
in the area surrounding a licensee's facility of the licensee's 
amendment and of its proposed determination as described in paragraph 
(a)(1) of this section, consulting with the licensee on the proposed 
media release and on the geographical area of its coverage;
    (ii) Will provide for a reasonable opportunity for the public to 
comment, using its best efforts to make available to the public 
whatever means of communication it can for the public to respond 
quickly, and, in the case of telephone comments, have these comments 
recorded or transcribed, as necessary and appropriate;
    (iii) When it has issued a local media release, may inform the 
licensee of the public's comments, as necessary and appropriate;
    (iv) Will publish a notice of issuance under Sec.  2.106 of this 
chapter;
    (v) Will provide a hearing after issuance, if one has been 
requested by a person who satisfies the provisions for intervention 
specified in Sec.  2.309 of this chapter; and
    (vi) Will require the licensee to explain the exigency and why the 
licensee cannot avoid it and use its normal public notice and comment 
procedures in paragraph (a)(1) of this section if it determines that 
the licensee has failed to use its best efforts to make a timely 
application for the amendment in order to create the exigency and to 
take advantage of this procedure.
    (6) Where the Commission finds that significant hazards 
considerations are involved, it will issue a Federal Register notice 
providing an opportunity for a prior hearing even in an emergency 
situation, unless it finds an imminent danger to the health or safety 
of the public, in which case it will issue an appropriate order or rule 
under 10 CFR part 2.
    (b) State consultation. (1) At the time a licensee requests an 
amendment, it must notify the State in which its facility is located of 
its request by providing that State with a copy of its application and 
its reasoned analysis about no significant hazards considerations and 
indicate on the application that it has done so.
    (2) The Commission will advise the State of its proposed 
determination about no significant hazards consideration normally by 
sending it a copy of the Federal Register notice.
    (3) The Commission will make the names of the Project Manager or 
other NRC personnel it designated to consult with the State available 
to the State official designated to consult about its proposed 
determination. The Commission will consider any comments of that State 
official. If it does not hear from the State in a timely manner, it 
will consider that the State has no interest in its determination; 
nonetheless, to ensure that the State is aware of the application, 
before it issues the amendment, it will make a good faith effort to 
communicate directly with that official. (Inability to consult with a 
responsible State official following good faith attempts will not 
prevent the Commission from making effective a license amendment 
involving no significant hazards consideration.)
    (4) The Commission will make a good faith attempt to consult with 
the State before it issues a license amendment involving no significant 
hazards consideration. If, however, it does not have time to use its 
normal consultation procedures because of an emergency situation, it 
will attempt to communicate directly with the appropriate State 
official. (Inability to consult with a responsible State official 
following good faith attempts will not prevent the Commission from 
making effective a license amendment involving no significant hazards 
consideration, if the Commission deems it necessary in an emergency 
situation.)
    (5) After the Commission issues the requested amendment, it will 
send a copy of its determination to the State.
    (c) Caveats about State consultation. (1) The State consultation 
procedures in paragraph (b) of this section do not give the State a 
right--
    (i) To veto the Commission's proposed or final determination;
    (ii) To a hearing on the determination before the amendment becomes 
effective; or
    (iii) To insist upon a postponement of the determination or upon 
issuance of the amendment.
    (2) These procedures do not alter present provisions of law that 
reserve to the Commission exclusive responsibility for setting and 
enforcing radiological

[[Page 87105]]

health and safety requirements for commercial nuclear plants.


Sec.  53.1520  Issuance of amendment.

    (a) In determining whether an amendment to a license will be issued 
to the applicant, the Commission will be guided by the considerations 
which govern the issuance of initial licenses to the extent applicable 
and appropriate. If the application is for amendment of an OL or COL 
and involves the material alteration of a commercial nuclear plant, a 
construction permit (CP) will be issued before the issuance of the 
amendment to the license, provided however, that if the application 
involves a material alteration to a manufactured reactor under this 
part before its installation at a site, or a COL before the date that 
the Commission makes the finding under Sec.  53.1452(g), no application 
for or issuance of a CP is required. If the amendment involves a 
significant hazards consideration, the Commission will give notice of 
its proposed action--
    (1) Under Sec.  2.105 of this chapter before acting thereon; and
    (2) As soon as practicable after the application has been docketed.
    (b) The Commission will be particularly sensitive to a license 
amendment request that involves irreversible consequences (such as one 
that permits a significant increase in the amount of effluents or 
radiation emitted by a commercial nuclear plant).
    (c) The Commission may make a final determination, under the 
procedures in Sec.  53.1515, that a proposed amendment to an OL or a 
COL for a commercial nuclear plant under this part involves no 
significant hazards consideration, if operation of the plant in 
accordance with the proposed amendment would not--
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated; or
    (2) Create the possibility of a new or different kind of an 
accident from any accident previously evaluated; or
    (3) Involve a significant reduction in a margin of safety.


Sec.  53.1525  Revising certification information within a design 
certification rule.

    (a) A holder of an OL or COL who references a design certification 
rule issued under this part must request an exemption if proposing to 
change one or more elements of the certification information. The 
Commission may grant such a request only if it determines that the 
exemption will comply with the requirements of Sec.  53.080 and that 
the special circumstances outweigh any decrease in safety that may 
result from the reduction in standardization caused by the departure.
    (b) The request for an exemption must be included with any 
associated license amendment request, which must be requested and 
processed in accordance with Sec. Sec.  53.1510, 53.1515, and 53.1520.
    (c) Licensees must evaluate changes to the design as described in 
the Final Safety Analysis Report (FSAR) not involving changes to the 
certification information using the criteria in Sec.  53.1550.


Sec.  53.1530  Revising design information within a manufacturing 
license.

    (a) The holder of an ML may not make changes to the design of the 
manufactured reactor authorized to be manufactured without obtaining an 
amendment pursuant to Sec.  53.1510 and, as applicable, Sec.  53.1520.
    (b) The holder of a COL under this part who references or uses a 
manufactured reactor under this part must request approval for any 
proposed departure from the design characteristics, site parameters, 
terms and conditions, or approved design of the manufactured reactor. 
The application for such departures must be submitted and processed in 
accordance with Sec. Sec.  53.1510, 53.1515, and 53.1520. In those 
cases where an ML references a design certification rule, the amendment 
application from the holder of the COL must also request an exemption 
from the design certification rule under Sec.  53.1525 if one or more 
elements of the certification information are adversely affected by the 
proposed change. The holder of the COL must evaluate changes to the 
commercial nuclear plant as described in the FSAR but outside of the 
scope of the referenced ML using the criteria in Sec.  53.1550.


Sec.  53.1535  Amendments during construction.

    (a) The holder of a CP or limited work authorization (LWA) under 
this part may request an amendment to the CP or LWA in order to gain 
Commission approval of the safety of selected design features or 
specifications, including proposed departures from a design 
certification rule or ML. Amendments to CPs or LWAs under this part 
must be requested and processed under Sec. Sec.  53.1510 and 53.1520.
    (b) The holder of a COL under this part for which the NRC has not 
yet made a finding in accordance with Sec.  53.1452(g) must request 
amendments required by Sec.  53.1525 or Sec.  53.1550 no later than 45 
days from the date the licensee begins the construction of the SSCs to 
implement the change or departure requiring NRC approval. The licensee 
proceeds with such changes at its own risk recognizing that there is a 
possibility that the amendment will not be granted.


Sec.  53.1540  Updating licensing-basis information and determining the 
need for NRC approval.

    (a) Sections 53.1545 through 53.1565 provide the process for a 
holder of an OL or COL to modify licensing-basis information and to 
evaluate potential changes to its facilities, procedures, programs, and 
organizations to determine if NRC approval is required.
    (b) Definitions for the purposes of Sec. Sec.  53.1545 through 
53.1565--
    Change means a modification or addition to, or removal from, the 
commercial nuclear plant or procedures that affects a design feature or 
related functional design criteria, method of performing or controlling 
the functions of design features, or an evaluation that demonstrates 
that intended functions will be accomplished.
    Departure from a method of evaluation described in the Final Safety 
Analysis Report (FSAR) (as updated) used in establishing the functional 
design criteria for safety-related structures, systems, or components 
or in the safety analyses means--
    (1) Changing any of the elements of the method described in the 
FSAR (as updated) unless the results of the analysis are conservative 
or essentially the same; or
    (2) Changing from a method described in the FSAR to another method 
unless that method has been approved by NRC for the intended 
application.
    Facility as described in the FSAR (as updated) means--
    (1) The SSCs that are described in the FSAR (as updated),
    (2) The design and performance requirements for such SSCs described 
in the FSAR (as updated), and
    (3) The evaluations or methods of evaluation included in the FSAR 
(as updated) for such SSCs which demonstrate that their intended 
function(s) will be accomplished.
    Final Safety Analysis Report (as updated) means the FSAR submitted 
under Sec.  53.1369 or Sec.  53.1416, as amended and supplemented, and 
as updated under Sec.  53.1545, as applicable.
    Procedures as described in the Final Safety Analysis Report (as 
updated) means those procedures that contain information described in 
the FSAR (as updated) such as how SSCs are operated and controlled 
(including assumed operator actions and response times).

[[Page 87106]]

Sec.  53.1545  Updating Final Safety Analysis Reports.

    (a) Each holder of an OL or COL under this part for which the 
Commission has made the finding under Sec.  53.1452(g) must update the 
FSAR originally submitted as part of the application for the license 
every 24 months or more frequently to assure that the information 
included in the report contains the latest information developed. The 
submittal must include the effects on the content of the FSAR of--
    (1) Changes made to the facility or procedures as described in the 
FSAR;
    (2) Safety analyses and evaluations performed by the licensee 
either in support of approved license amendments or in support of 
conclusions that changes did not require a license amendment under 
Sec.  53.1550;
    (3) Updates to the probabilistic risk assessments required under 
Sec.  53.450;
    (4) The cumulative effects of the changes to the facility or 
procedures on the margins to the safety criteria in Sec. Sec.  53.210, 
53.220, 53.450(e), and 53.470 since the last FSAR update; and
    (5) Analyses of new safety issues performed by or on behalf of the 
licensee at Commission request.
    (b)(1) The licensee must submit revisions containing updated 
information to the Commission, under Sec.  53.040, identifying the 
location of revised or new information.
    (2) The submittal must include--
    (i) A certification by a duly authorized officer of the licensee 
that either the information accurately presents changes made since the 
previous submittal, necessary to reflect information and analyses 
submitted to the Commission or prepared pursuant to Commission 
requirement, or that no such changes were made; and
    (ii) An identification of changes made under the provisions of 
Sec.  53.1550 but not previously submitted to the Commission.
    (c) Each applicant for or holder of a COL under this part for which 
the Commission has not made the finding under Sec.  53.1452(g) must 
submit an update to the FSAR annually by providing the information 
required in (a)(1) through (a)(5) of this section and meeting the 
requirements of paragraph (b) of this section. Combined license 
applicants who have requested the NRC to suspend its review of the COL 
application and COL holders who have informed the NRC that they do not 
plan to pursue construction need not submit an annual update of the 
FSAR. If a COL applicant requests that the NRC resume its review, or a 
COL holder notifies the NRC that the COL holder plans to commence or 
resume construction, then the COL applicant or holder must submit to 
NRC an update to its FSAR within 90 days of the request or 
notification, as applicable, and annually thereafter.
    (d) The FSAR (as updated) must be retained by the licensee until 
the Commission terminates its license.
    (e) Each holder of an ML under this part must submit an update of 
the FSAR reflecting any modification to the design that is directed or 
approved by the Commission under Sec.  53.1288 or Sec.  53.1530, and 
any new analyses of the design requested by the Commission under Sec.  
53.1580.


Sec.  53.1550  Evaluating changes to facility as described in Final 
Safety Analysis Reports.

    (a) The holder of an OL or COL may make changes in the facility as 
described in the FSAR (as updated) and make changes in the procedures 
as described in the FSAR (as updated) without obtaining a license 
amendment pursuant to Sec.  53.1510 only if--
    (1) A change to the technical specifications incorporated in the 
license is not required; and
    (2) The change meets all of the following criteria:
    (i) Does not result in an increase to the frequency or consequences 
of an event sequence such that an event sequence not previously 
identified as risk significant becomes risk significant by the analyses 
performed in accordance with Sec.  53.450(e).
    (ii) Does not result in an increase to the frequency or 
consequences of an event sequence such that an event sequence 
identified as risk significant in accordance with Sec.  53.450(e) 
exceeds the licensing-basis event evaluation criteria required to be 
established in accordance with Sec.  53.450(e).
    (iii) Does not involve either of the following: (A) a change to the 
NRC-approved comprehensive risk metric(s) or associated risk 
performance objective under Sec.  53.220(b), or (B) an increase to the 
frequency or consequences of one or more event sequences such that 
there is more than a minimal reduction in the margin between the 
calculated comprehensive risks posed by the commercial nuclear plant 
and the safety criteria of Sec.  53.220.
    (iv) Does not involve a departure from a method of evaluation 
described in the FSAR (as updated) used in assessing licensing-basis 
events in accordance with Sec.  53.450 unless the results of the 
analysis under Sec.  53.450 are conservative or essentially the same, 
the revised method of evaluation has been previously approved by the 
NRC for the intended application, or the revised method of evaluation 
can be used under an NRC-endorsed consensus code or standard.
    (v) Does not result in the escalation in the safety classification 
of an SSC from non-safety-related to non-safety-related but safety-
significant or from non-safety-related but safety-significant to 
safety-related.
    (vi) Does not result in more than a minimal decrease in defense in 
depth.
    (vii) For commercial nuclear plants licensed under this part for 
which alternative evaluation criteria are adopted in accordance with 
Sec.  53.470, does not result in a change to the frequency or 
consequences of event sequences such that the calculated margins 
between the results for event sequences evaluated in accordance with 
Sec.  53.450(e) and the alternative evaluation criteria decreases by 25 
percent or more.
    (viii) Does not result in the identification of a new design-basis 
accident in accordance with Sec.  53.450(f).
    (ix) Does not result in a decrease by 10 percent or more in the 
margin between the consequence of any design-basis accident and the 
safety criteria in Sec.  53.210.
    (x) Does not prevent meeting the design requirements in Sec.  
53.440(j) to limit the release of radionuclides from reactor systems, 
waste stores, or other significant inventories of radioactive materials 
assuming the impact of a large, commercial aircraft.
    (3) In implementing this paragraph, the FSAR (as updated) is 
considered to include FSAR changes since submittal of the last update 
of the FSAR under Sec.  53.1545.
    (4) The provisions in this section do not apply to changes to the 
facility or procedures when the applicable regulations establish more 
specific criteria for accomplishing such changes.
    (b)(1) A licensee who references a design certification rule may 
make departures from the standard design, without prior Commission 
approval, unless the proposed departure involves a change to the design 
as described in the rule certifying the design, in which case the 
requirements of Sec.  53.1525 are applicable.
    (2) The licensee must maintain records of all departures from the 
certified design of the facility and these records must be maintained 
and available for audit until the termination of the license. The 
licensee must identify the location and nature of departures from 
licensing-basis information within supporting documents for a certified 
design within the updates to the Safety Analysis Report required by 
Sec.  53.1545.

[[Page 87107]]

    (3) Licensees for which the NRC has docketed the certifications 
required under Sec.  53.1070 need not retain records of departures from 
the design of the facility associated with SSCs that have been 
permanently removed from service using an NRC-approved change process.
    (c)(1) The licensee must maintain records of changes in the 
facility and procedures made under paragraph (a) of this section. These 
records must include a written evaluation which provides the bases for 
the determination that the change does not require a license amendment 
under paragraph (a)(2) of this section.
    (2) The licensee must submit, as specified in Sec.  53.040, a 
report containing a brief description of any departures and changes, 
including a summary of the evaluation of each. A report must be 
submitted at intervals not to exceed 24 months. For COLs, the report 
must be submitted at intervals not to exceed 6 months during the period 
from the date of application for a COL to the date the Commission makes 
its findings under Sec.  53.1452(g).
    (3) The records of changes in the facility must be maintained until 
the termination of an OL or COL issued under this part, or the 
termination of a renewed license issued under Sec.  53.1595--whichever 
is later. Records of changes in procedures must be maintained for a 
period of 5 years.


Sec.  53.1560  Updating program documents included in licensing-basis 
information.

    (a) Each holder under this part of an OL or COL for which the 
Commission has made the finding under Sec.  53.1452(g) must biennially 
or more frequently update the program documents submitted as part of an 
application to obtain or maintain the license to assure that the 
information included in the documents contains the latest information 
developed. The submittals must include the effects on the content of 
the program documents of--
    (1) Changes made in the facility, procedures, licensee's 
organization, or site environs;
    (2) Safety analyses and evaluations performed by the applicant or 
licensee either in support of approved license amendments or in support 
of conclusions that changes did not require a license amendment in 
accordance with Sec.  53.1550;
    (3) Analyses of new safety issues performed by or on behalf of the 
licensee at Commission request; and
    (4) Changes to the programs as a result of operating experience, 
corrective actions, or other reasons deemed appropriate to ensure the 
programs serve their underlying purpose to support the requirements in 
subpart B of this part or other NRC regulations.
    (b)(1) The licensee must submit revisions containing updated 
information to the Commission, as specified in Sec.  53.040, 
identifying the location of revised or new information.
    (2) The submittal must include--
    (i) A certification by a duly authorized officer of the licensee 
that either the information accurately presents changes made since the 
previous submittals, necessary to reflect information and analyses 
submitted to the Commission or prepared pursuant to Commission 
requirement, or that no such changes were made; and
    (ii) An identification of changes made under the provisions of 
Sec.  53.1550 but not previously submitted to the Commission.
    (c) The updated program documents must be retained by the licensee 
until the Commission terminates their license.


Sec.  53.1565  Evaluating changes to programs included in licensing-
basis information.

    (a) A licensee may make changes to the facility, procedures, or 
organizations or address changes to site environs as described in the 
program documents included in licensing-basis information without 
obtaining prior NRC approval only if--
    (1) A change to the technical specifications incorporated in the 
license is not required;
    (2) An exemption from an NRC regulation is not required; and
    (3) The change conforms to program-specific requirements included 
in regulations in this part, technical specifications, or the NRC-
approved program document included and reviewed as part of a license 
application under subpart H or an amendment under this subpart.
    (b) In implementing this section, the program documents (as 
updated) include changes since submittal of the last updates of the 
program documents pursuant to Sec.  53.1560.
    (c) The provisions in this section do not apply to changes to the 
program documents when the applicable regulations establish more 
specific criteria for accomplishing such changes.
    (d) To make changes to the facility, procedures, or organizations 
or to address changes to site environs as described in the program 
documents included in licensing-basis information for individual 
programs, the following requirements must be satisfied:
    (1) Quality assurance program--operation. (i) Each holder under 
this part of an OL or COL, after the Commission makes the finding under 
Sec.  53.1452(g), may make a change to a previously accepted quality 
assurance program (QAP) description included or referenced in the 
Safety Analysis Report without prior NRC approval, provided the change 
does not reduce the commitments in the program description as accepted 
by the NRC. Changes to the QAP description that do not reduce the 
commitments must be submitted to the NRC in accordance with the 
requirements of Sec.  53.1545. In addition to QAP changes involving 
administrative improvements and clarifications, spelling corrections, 
punctuation, or editorial items, the following changes are not 
considered to be reductions in commitment:
    (A) The use of a quality assurance (QA) standard approved by the 
NRC which is more recent than the QA standard in the licensee's QAP at 
the time of the change;
    (B) The use of a QA alternative or exception approved by an NRC 
safety evaluation, provided that the bases of the NRC approval are 
applicable to the licensee's facility;
    (C) The use of generic organizational position titles that clearly 
denote the position function, supplemented as necessary by descriptive 
text, rather than specific titles;
    (D) The use of generic organizational charts to indicate functional 
relationships, authorities, and responsibilities, or, alternately, the 
use of descriptive text;
    (E) The elimination of QAP information that duplicates language in 
QA regulatory guides and QA standards to which the licensee is 
committed; and
    (F) Organizational revisions that ensure that persons and 
organizations performing QA functions continue to have the requisite 
authority and organizational freedom, including sufficient independence 
from cost and schedule when opposed to safety considerations.
    (ii) Changes to the QAP description that do reduce the commitments 
must be submitted to the NRC and receive NRC approval prior to 
implementation, as follows:
    (A) Changes made to the QAP description as presented in the Safety 
Analysis Report or in a topical report must be submitted as specified 
in Sec.  53.040.
    (B) The submittal of a change to the Safety Analysis Report QAP 
description must include all pages affected by that change and must be 
accompanied by a forwarding letter identifying the change, the reason 
for the change, and the basis for concluding that the revised program 
incorporating the change continues to satisfy the criteria of appendix 
B to part

[[Page 87108]]

50 of this chapter and the Safety Analysis Report QAP description 
commitments previously accepted by the NRC (the letter need not provide 
the basis for changes that correct spelling, punctuation, or editorial 
items).
    (C) A copy of the forwarding letter identifying the change must be 
maintained as a facility record for 3 years.
    (D) Changes to the QAP description included or referenced in the 
Safety Analysis Report shall be regarded as accepted by the Commission 
upon receipt of a letter to this effect from the appropriate reviewing 
office of the Commission or 60 days after submittal to the Commission, 
whichever occurs first.
    (2) Quality assurance program--siting, construction, and 
manufacturing. Each holder of an LWA, early site permit, CP, ML, or 
COL, before the Commission makes the finding under Sec.  53.1452(g) of 
this chapter, under this part may make a change to a previously 
accepted QAP description included or referenced in the Safety Analysis 
Report without prior NRC approval, provided the change does not reduce 
the commitments in the program description previously accepted by the 
NRC. Changes to the QAP description that do not reduce the commitments 
must be submitted to NRC within 90 days. Changes to the QAP description 
that reduce the commitments must be submitted to NRC and receive NRC 
approval before implementation, as follows:
    (i) Changes to the Safety Analysis Report must be submitted for 
review as specified in Sec.  53.040. Changes made to NRC-accepted QA 
topical report descriptions must be submitted as specified in Sec.  
53.040.
    (ii) The submittal of a change to the Safety Analysis Report QAP 
description must include all pages affected by that change and must be 
accompanied by a forwarding letter identifying the change, the reason 
for the change, and the basis for concluding that the revised program 
incorporating the change continues to satisfy the criteria of appendix 
B of part 50 of this chapter and the Safety Analysis Report QAP 
description commitments previously accepted by the NRC (the letter need 
not provide the basis for changes that correct spelling, punctuation, 
or editorial items).
    (iii) A copy of the forwarding letter identifying the changes must 
be maintained as a facility record for 3 years.
    (iv) Changes to the QAP description included or referenced in the 
Safety Analysis Report shall be regarded as accepted by the Commission 
upon receipt of a letter to this effect from the appropriate reviewing 
office of the Commission or 60 days after submittal to the Commission, 
whichever occurs first.
    (3) Emergency preparedness program. (i) Definitions for the purpose 
of paragraph (d)(3) of this section:
    (A) Change means an action that results in modification or addition 
to, or removal from, the licensee's emergency plan. All such changes 
are subject to the provisions of this section except where the 
applicable regulations establish specific criteria for accomplishing a 
particular change.
    (B) Emergency plan means the document(s), prepared and maintained 
by the licensee, that identify and describe the licensee's methods for 
maintaining emergency preparedness and responding to emergencies. An 
emergency plan includes the plan as originally approved by the NRC and 
all subsequent changes made by the licensee with, and without, prior 
NRC review and approval under paragraph (d)(3) of this section.
    (C) Emergency planning function means a capability or resource 
necessary to prepare for and respond to a radiological emergency.
    (D) Reduction in effectiveness means a change in an emergency plan 
that results in reducing the licensee's capability to perform an 
emergency planning function in the event of a radiological emergency.
    (ii)(A) Except as provided in paragraph (d)(3)(ii)(B) of this 
section, a holder of an OL under this part, or a COL under this part 
after the Commission makes the finding under Sec.  53.1452(g), must 
follow and maintain the effectiveness of an emergency plan that meets 
the requirements in appendix E to part 50 of this chapter and the 
planning standards of Sec.  50.47(b).
    (B) A holder of an OL under this part for a commercial nuclear 
plant consisting of small modular reactors (SMRs) or non-light-water 
reactors, or a holder of a COL under this part after the Commission 
makes the finding under Sec.  53.1452(g) for a commercial nuclear plant 
consisting of either SMRs or non-light-water reactors, must follow and 
maintain the effectiveness of either an emergency plan that meets the 
requirements in Sec.  50.160 or an emergency plan that meets the 
requirements in appendix E to part 50 of this chapter and the planning 
standards of Sec.  50.47(b).
    (iii)(A) Except as provided in paragraph (d)(3)(iii)(B) of this 
section, the licensee may make changes to its emergency plan without 
NRC approval only if the licensee performs and retains an analysis 
demonstrating that the changes do not reduce the effectiveness of the 
plan and the plan, as changed, continues to meet the requirements in 
appendix E to part 50 of this chapter and the planning standards of 
Sec.  50.47(b).
    (B) A license under this part for a commercial nuclear plant 
consisting of either SMRs or non-light-water reactors may make changes 
to its emergency plan without NRC approval only if the licensee 
performs and retains an analysis demonstrating that the changes do not 
reduce the effectiveness of the plan and the plan, as changed, 
continues to meet either the requirements in Sec.  50.160 or the 
requirements in appendix E to part 50 and the planning standards of 
Sec.  50.47(b).
    (iv) The changes to a licensee's emergency plan that reduce the 
effectiveness of the plan as defined in paragraph (d)(3)(i)(D) of this 
section may not be implemented without prior approval by the NRC. A 
licensee desiring to make such a change must submit an application for 
an amendment to its license. In addition to the filing requirements of 
Sec. Sec.  53.1510 and 53.1515, the request must include all emergency 
plan pages affected by that change and must be accompanied by a 
forwarding letter identifying the change, the reason for the change, 
and the basis for concluding that the licensee's emergency plan, as 
revised, will continue to meet either the requirements in Sec.  50.160 
to this chapter or the requirements in appendix E to part 50 of this 
chapter and the planning standards of Sec.  50.47(b) of this chapter.
    (v) The licensee must retain a record of each change to the 
emergency plan made without prior NRC approval for a period of three 
years from the date of the change and shall submit, as specified in 
Sec.  53.040, a report of each such change, including a summary of its 
analysis, within 30 days after the change is put in effect.
    (vi) The licensee must retain the emergency plan and each change 
for which prior NRC approval was obtained pursuant to paragraph 
(d)(3)(iv) of this section as a record until the Commission terminates 
the license for the nuclear power reactor.
    (vii)(A) The licensee must provide for the development, revision, 
implementation, and maintenance of its emergency preparedness program. 
The licensee must ensure that all program elements are reviewed by 
persons who have no direct responsibility for the implementation of the 
emergency preparedness program either--

[[Page 87109]]

    (1) At intervals not to exceed 12 months; or
    (2) As necessary, based on an assessment by the licensee against 
performance indicators, and as soon as reasonably practicable after a 
change occurs in personnel, procedures, equipment, or facilities that 
potentially could adversely affect emergency preparedness, but no 
longer than 12 months after the change. In any case, all elements of 
the emergency preparedness program must be reviewed at least once every 
24 months.
    (B) The review must include an evaluation for adequacy of 
interfaces with State participating Tribal and local governments and of 
licensee drills, exercises, capabilities, and procedures. The results 
of the review, along with recommendations for improvements, must be 
documented, reported to the licensee's corporate and plant management, 
and retained for a period of 5 years. The part of the review involving 
the evaluation for adequacy of interface with State, participating 
Tribal and local governments must be available to the appropriate 
State, participating Tribal and local governments.
    (4) Security programs. (i) The licensee must prepare and maintain 
safeguards contingency plan procedures in accordance with appendix C of 
part 73 of this chapter for affecting the actions and decisions 
contained in the Responsibility Matrix of the safeguards contingency 
plan. The licensee may not make a change that would decrease the 
safeguard effectiveness of a physical security plan, or guard training 
and qualification plan, or cybersecurity plan submitted under subpart H 
or part 73 of this chapter, or of the first four categories of 
information (Background, Generic Planning Base, Licensee Planning Base, 
Responsibility Matrix) contained in a licensee safeguards contingency 
plan submitted under subpart H or part 73 of this chapter, as 
applicable, without prior approval of the Commission. A licensee 
desiring to make such a change must submit an application for amendment 
to the licensee's license under Sec. Sec.  53.1510, 53.1515, and 
53.1520.
    (ii) The licensee may make changes to the plans referenced in 
paragraph (4)(i) of this section without prior Commission approval if 
the changes do not decrease the safeguards effectiveness of the plan. 
The licensee must maintain records of changes to the plans made without 
prior Commission approval for a period of 3 years from the date of the 
change, and must submit, as specified in Sec.  53.040, a report 
containing a description of each change within 2 months after the 
change is made. Prior to the safeguards contingency plan being put into 
effect, the licensee must have--
    (A) All safeguards capabilities specified in the safeguards 
contingency plan available and functional;
    (B) Detailed procedures developed according to appendix C to part 
73 of this chapter available at the licensee's site; and
    (C) All appropriate personnel trained to respond to safeguards 
incidents as outlined in the plan and specified in the detailed 
procedures.
    (iii) The licensee must provide for the development, revision, 
implementation, and maintenance of its safeguards contingency plan. The 
licensee must ensure that all program elements are reviewed by 
individuals independent of both security program management and 
personnel who have direct responsibility for implementation of the 
security program either--
    (A) At intervals not to exceed 12 months; or
    (B) As necessary, based on an assessment by the licensee against 
performance indicators, and as soon as reasonably practicable after a 
change occurs in personnel, procedures, equipment, or facilities that 
potentially could adversely affect security, but no longer than 12 
months after the change. In any case, all elements of the safeguards 
contingency plan must be reviewed at least once every 24 months.
    (iv) The review must include a review and audit of safeguards 
contingency procedures and practices, an audit of the security system 
testing and maintenance program, and a test of the safeguards systems 
along with commitments established for response by local law 
enforcement authorities. The results of the review and audit, along 
with recommendations for improvements, must be documented, reported to 
the licensee's corporate and plant management, and kept available at 
the plant for inspection for a period of 3 years.


Sec.  53.1570  Transfer of licenses.

    (a) No commercial nuclear plant license issued under this part, or 
any right thereunder, shall be transferred, assigned, or in any manner 
disposed of, either voluntarily or involuntarily, directly or 
indirectly, through transfer of control of the license to any person, 
unless the Commission gives its consent in writing.
    (b)(1) An application for transfer of a license must include--
    (i) As much of the information described in Sec. Sec.  53.1109, 
53.1306, 53.1366, and 53.1413 with respect to the identity and 
technical and financial qualifications of the proposed transferee as 
would be required by those sections if the application were for an 
initial license. The Commission may require additional information such 
as data respecting proposed safeguards against hazards from radioactive 
materials and the applicant's qualifications to protect against such 
hazards.
    (ii) A statement of the purposes for which the transfer of the 
license is requested, the nature of the transaction necessitating or 
making desirable the transfer of the license, and an agreement to limit 
access to Restricted Data or Classified National Security Information 
pursuant to Sec.  53.1115. The Commission may require any person who 
submits an application for license pursuant to the provisions of this 
section to file a written consent from the existing licensee or a 
certified copy of an order or judgment of a court of competent 
jurisdiction attesting to the person's right (subject to the licensing 
requirements of the Act and these regulations) to possession of the 
facility or site involved.
    (2) [Reserved]
    (c) After appropriate notice to interested persons, including the 
existing licensee, and observance of such procedures as may be required 
by the Act or regulations or orders of the Commission, the Commission 
will approve an application for the transfer of a license, if the 
Commission determines--
    (1) That the proposed transferee is qualified to be the holder of 
the license; and
    (2) That transfer of the license is otherwise consistent with 
applicable provisions of law, regulations, and orders issued by the 
Commission pursuant thereto.


Sec.  53.1575  Termination of licenses.

    (a) When the holder of an OL or COL under this part has determined 
to permanently cease operations the licensee must, within 30 days, 
submit a written certification to the NRC, consistent with the 
requirements of Sec.  53.1070.
    (b) Once fuel has been permanently removed from the reactor system, 
the licensee must submit a written certification to the NRC that meets 
the requirements of Sec.  53.1070.
    (c)(1) Upon docketing of the certifications for permanent cessation 
of operations and permanent removal of fuel from the reactor system, or 
when a final legally effective order to permanently cease operations 
has come into effect, the license no longer authorizes operation of the 
reactor or

[[Page 87110]]

emplacement or retention of fuel into the reactor system.
    (2) Activities associated with decommissioning will be carried out 
in accordance with the requirements and procedures in subpart G of this 
part.
    (3) The Commission shall terminate the license if it determines 
that--
    (i) The remaining dismantlement has been performed in accordance 
with the approved license termination plan required in subpart G of 
this part; and
    (ii) The final radiation survey and associated documentation, 
including an assessment of dose contributions associated with parts 
released for use before approval of the license termination plan, 
demonstrate that the facility and site have met the criteria for 
decommissioning in subpart E of 10 CFR part 20.
    (d) A holder of a CP or COL under this part may request the 
termination of the license as well as licenses issued by the NRC under 
parts 30, 40, or 70 of this chapter prior to plant operations. Such 
requests may support an immediate NRC approval of the site for 
unrestricted use.


Sec.  53.1580  Information requests.

    Each licensee under this part must at any time before termination 
of the license, upon request of the Commission, submit, as specified in 
Sec.  53.040 written statements, signed under oath or affirmation, to 
enable the Commission to determine whether or not the license should be 
modified, suspended, or revoked. Except for information sought to 
verify licensee compliance with the current licensing basis for that 
facility, the NRC must prepare the reason or reasons for each 
information request prior to issuance to ensure that the burden to be 
imposed on respondents is justified in view of the potential safety 
significance of the issue to be addressed in the requested information. 
Each such justification provided for an evaluation performed by the NRC 
staff must be approved by the Executive Director for Operations or his 
or her designee prior to issuance of the request.


Sec.  53.1585  Revocation, suspension, modification of licenses and 
approvals for cause.

    A license or standard design approval issued under this part may be 
revoked, suspended, or modified, in whole or in part, for any material 
false statement in the application or in the supplemental or other 
statement of fact required of the applicant; or because of conditions 
revealed by the application or statement of fact of any report, record, 
inspection, or other means which would warrant the Commission to refuse 
to grant a license or approval on an original application; or for 
failure to manufacture a reactor, or construct or operate a facility in 
accordance with the terms of the license, provided, however, that 
failure to make timely completion of the proposed construction or 
alteration of a facility under a CP under this part shall be governed 
by the provisions of Sec.  53.1342(b); or for violation of, or failure 
to observe, any of the terms and provisions of the Act, regulations, 
license, approval, or order of the Commission.


Sec.  53.1590  Backfitting.

    (a)(1) Backfitting means the modification of or addition to 
systems, structures, components, or design of a facility; or the design 
approval or ML for a facility; or the procedures or organization 
required to design, construct or operate a facility; any of which may 
result from a new or amended provision in the Commission's regulations 
or the imposition of a regulatory staff position interpreting the 
Commission's regulations that is either new or different from a 
previously applicable staff position after the date of the commercial 
nuclear plant license issued under this part.
    (2) Except as provided in paragraph (a)(4) of this section, the 
Commission shall require a systematic and documented analysis pursuant 
to paragraph (b) of this section for backfits which it seeks to impose.
    (3) Except as provided in paragraph (a)(4) of this section, the 
Commission shall require the backfitting of a facility only when it 
determines, based on the analysis described in paragraph (b) of this 
section, that there is a substantial increase in the overall protection 
of the public health and safety or the common defense and security to 
be derived from the backfit and that the direct and indirect costs of 
implementation for that facility are justified in view of this 
increased protection.
    (4) The provisions of paragraphs (a)(2) and (a)(3) of this section 
are inapplicable and, therefore, backfit analysis is not required and 
the standards in paragraph (a)(3) of this section do not apply where 
the Commission or staff, as appropriate, finds and declares, with 
appropriate documented evaluation for its finding, either--
    (i) That a modification is necessary to bring a facility into 
compliance with a license or the rules or orders of the Commission, or 
into conformance with written commitments by the licensee; or
    (ii) That regulatory action is necessary to ensure that the 
facility provides adequate protection to the health and safety of the 
public and is in accord with the common defense and security; or
    (iii) That the regulatory action involves defining or redefining 
what level of protection to the public health and safety or common 
defense and security should be regarded as adequate.
    (5) The Commission must always require the backfitting of a 
facility if it determines that such regulatory action is necessary to 
ensure that the facility provides adequate protection to the health and 
safety of the public and is in accord with the common defense and 
security.
    (6) The documented evaluation required by paragraph (a)(4) of this 
section must include a statement of the objectives of and reasons for 
the modification and the basis for invoking the exception. If 
immediately effective regulatory action is required, then the 
documented evaluation may follow rather than precede the regulatory 
action.
    (7) If there are two or more ways to achieve compliance with a 
license or the rules or orders of the Commission, or with written 
licensee commitments, or there are two or more ways to reach a level of 
protection which is adequate, then ordinarily the applicant or licensee 
is free to choose the way which best suits its purposes. However, 
should it be necessary or appropriate for the Commission to prescribe a 
specific way to comply with its requirements or to achieve adequate 
protection, then cost may be a factor in selecting the way, provided 
that the objective of compliance or adequate protection is met.
    (b) In reaching the determination required by paragraph (a)(3) of 
this section, the Commission will consider how the backfit should be 
scheduled in light of other ongoing regulatory activities at the 
facility and, in addition, will consider information available 
concerning any of the following factors as may be appropriate and any 
other information relevant and material to the proposed backfit:
    (1) The statement of the specific objectives that the proposed 
backfit is designed to achieve;
    (2) The general description of the activity that would be required 
by the licensee or applicant in order to complete the backfit;
    (3) The potential change in the risk to the public from the 
accidental off-site release of radioactive material;
    (4) The potential impact on radiological exposure of facility 
employees;

[[Page 87111]]

    (5) The installation and continuing costs associated with the 
backfit, including the cost of facility downtime or the cost of 
construction delay;
    (6) The potential safety impact of changes in plant or operational 
complexity, including the relationship to proposed and existing 
regulatory requirements;
    (7) The estimated resource burden on the NRC associated with the 
proposed backfit and the availability of such resources;
    (8) The potential impact of differences in facility type, design or 
age on the relevancy and practicality of the proposed backfit;
    (9) Whether the proposed backfit is interim or final and, if 
interim, the justification for imposing the proposed backfit on an 
interim basis.
    (c) No licensing action will be withheld during the pendency of 
backfit analyses required by the Commission's rules.
    (d) The Executive Director for Operations shall be responsible for 
implementation of this section, and all analyses required by this 
section shall be approved by the Executive Director for Operations or 
his or her designee.


Sec.  53.1595  Renewal.

    Licenses may be renewed by the Commission upon expiration of the 
period of the license.

Subpart J--Reporting and Other Administrative Requirements


Sec.  53.1600  General information.

    Each applicant and licensee under this part must ensure that U.S. 
Nuclear Regulatory Commission (NRC) inspectors have unfettered access 
to sites and facilities licensed or proposed to be licensed in Sec.  
53.1610, must maintain records and make reports to the NRC in 
accordance with requirements in Sec. Sec.  53.1620 through 53.1650, 
must satisfy financial qualification and reporting requirements in 
Sec. Sec.  53.1660 through 53.1700, and must obtain and maintain 
required financial protections in case of an accident in Sec. Sec.  
53.1720 and 53.1730.


Sec.  53.1610  Unfettered access for inspections.

    (a) Each applicant for or holder of a manufacturing license (ML), 
operating license (OL), combined license (COL), construction permit 
(CP), or early site permit must permit inspection, by duly authorized 
representatives of the Commission, of its records, premises, 
activities, and of licensed materials in possession or use, related to 
the license or CP or early site permit as may be necessary to 
effectuate the purposes of the Atomic Energy Act of 1956, as amended, 
(the Act) and the Energy Reorganization Act of 1974, as amended.
    (b)(1) Each holder of an ML, OL, COL, or CP must, upon request by 
the Director, Office of Nuclear Reactor Regulation, provide rent-free 
office space for the exclusive use of the Commission inspection 
personnel. Heat, air conditioning, light, electrical outlets, and 
janitorial services must be furnished by each licensee and each holder 
of a CP. The office must be convenient to and have full access to the 
facility and must provide the inspectors both visual and acoustic 
privacy.
    (2) For a site or facility with an assigned resident inspector, the 
space provided must be adequate to accommodate a full-time inspector, a 
part-time secretary, and transient NRC personnel and must be generally 
commensurate with other office facilities at the site. For sites or 
facilities assigned multiple resident inspectors, additional space may 
be requested. The office space that is provided must be subject to the 
approval of the Director, Office of Nuclear Reactor Regulation. All 
furniture, supplies, and communication equipment will be furnished by 
the Commission.
    (3) For a site or facility without an assigned resident inspector, 
temporary space to accommodate periodic or special inspections must be 
provided. The office space must be generally commensurate with other 
office accommodations at the site.
    (4) The licensee or permit holder must afford any NRC resident 
inspector assigned to that site, or other NRC inspectors identified by 
the Regional Administrator as likely to inspect the facility, immediate 
unfettered access, equivalent to access provided regular plant 
employees, following proper identification and compliance with 
applicable access control measures for security, radiological 
protection, and personal safety.
    (5) The licensee or permit holder must ensure that the arrival and 
presence of an NRC inspector, who has been properly authorized facility 
access as described in paragraph (b)(4) of this section, is not 
announced or otherwise communicated by its employees or contractors to 
other persons at the facility unless specifically requested by the NRC 
inspector.


Sec.  53.1620  Maintenance of records, making of reports.

    (a) Each holder of an ML, OL, COL, CP, or early site permit must 
maintain all records and make all reports, in connection with the 
activity, as may be required by the conditions of the license or permit 
or by the regulations and orders of the Commission in effectuating the 
purposes of the Act and the Energy Reorganization Act of 1974, as 
amended. Reports must be submitted in accordance with Sec.  53.040.
    (b) [Reserved]
    (c) Records that are required by the regulations in this part, by 
license condition, or by technical specifications must be retained for 
the period specified by the appropriate regulation, license condition, 
or technical specification. If a retention period is not otherwise 
specified, these records must be retained until the Commission 
terminates the facility license or, in the case of an early site 
permit, until the permit expires.
    (d)(1) Records which must be retained under this part may be the 
original or a reproduced copy or a microform if the reproduced copy or 
microform is duly authenticated by authorized personnel and the 
microform is capable of producing a clear and legible copy after 
storage for the period specified by Commission regulations. The record 
may also be stored in electronic media with the capability of producing 
legible, accurate, and complete records during the required retention 
period. Records such as letters, drawings, and specifications, must 
include all pertinent information such as stamps, initials, and 
signatures. The licensee must maintain adequate safeguards against 
tampering with, and loss of records.
    (2) If there is a conflict between the Commission's regulations in 
this part, license condition, or technical specification, or other 
written Commission approval or authorization pertaining to the 
retention period for the same type of record, the retention period 
specified in the regulations in this part for such records shall apply 
unless the Commission, under Sec.  53.080 of this part, has granted a 
specific exemption from the record retention requirements in the 
regulations in this part.
    (e) Each licensee must notify the Commission as specified in Sec.  
53.040 of this part, of successfully completing power ascension testing 
or startup testing as applicable within 30 calendar days of completing 
the testing.


Sec.  53.1630  Immediate notification requirements for operating 
commercial nuclear plants.

    (a) General requirements.\1\ (1) Each holder of an OL under this 
part or a COL under this part after the Commission makes the finding 
under Sec.  53.1452(g), must notify the NRC Operations Center

[[Page 87112]]

via the Emergency Notification System (ENS) of--
    (i) The declaration of any of the Emergency Classes specified in 
the licensee's approved Emergency Plan; or
    (ii) Those non-emergency events specified in paragraph (b) of this 
section that occurred within 3 years of the date of discovery.
    (2) If the ENS is inoperative, the licensee must make the required 
notifications via commercial telephone service, other dedicated 
telephone system, or any other method which will ensure that a report 
is made as soon as practical to the NRC Headquarters Operations Center 
at the numbers specified in appendix A to part 73 of this chapter.
    (3) The licensee must notify the NRC immediately after notification 
of the appropriate State or local agencies and not later than 1 hour 
after the time the licensee declares one of the Emergency Classes.
    (4) The licensee must activate the data links with the NRC as 
specified in their emergency plans after declaring an Emergency Class 
for events of actual or potential substantial degradation of plant 
safety or security, probable risk to site personnel life, or site 
equipment damage caused by hostile action. The data links may also be 
activated by the licensee during emergency drills or exercises if the 
licensee's computer system has the capability to transmit the exercise 
data.
    (5) When making a report under paragraph (a)(1) of this section, 
the licensee must identify--
    (i) The Emergency Class declared; or
    (ii) Paragraph (b)(1), ``One-hour reports,'' paragraph (b)(2), 
``Four-hour reports,'' or paragraph (b)(3), ``Eight-hour reports,'' as 
the paragraph of this section requiring notification of the non-
emergency event.
    (b) Non-emergency events. (1) One-hour reports. If not reported as 
a declaration of an Emergency Class under paragraph (a) of this 
section, the licensee must notify the NRC as soon as practical and in 
all cases within one hour of the occurrence of any deviation from the 
plant's Technical Specifications authorized under Sec.  53.740(h) of 
this part.
    (2) Four-hour reports. If not reported under paragraphs (a) or 
(b)(1) of this section, the licensee must notify the NRC as soon as 
practical, and in all cases, within 4 hours of the occurrence of any of 
the following:
    (i) The initiation of any commercial nuclear plant shutdown 
required by the plant's Technical Specifications.
    (ii) Any event or condition that results in actuation of the 
reactor protection system when the reactor is critical except when the 
actuation results from and is part of a pre-planned sequence during 
testing or reactor operation.
    (iii) Any event or condition that results in an unplanned actuation 
of a safety-related (SR) standby cooling system or the unplanned sole 
reliance on an SR standby cooling system for those systems that are in 
constant operation.
    (iv) Any event or condition that results in an unplanned movement 
of, change of state in, or chemical interaction involving a significant 
amount of radioactive material within the commercial nuclear plant.
    (v) Any event or situation, related to the health and safety of the 
public or onsite personnel, or protection of the environment, for which 
a news release is planned or notification to other government agencies 
has been or will be made. Such an event may include an onsite fatality 
or inadvertent release of radioactively contaminated materials.
    (3) Eight-hour reports. If not reported under paragraphs (a), 
(b)(1), or (b)(2) of this section, the licensee must notify the NRC as 
soon as practical and in all cases within 8 hours of the occurrence of 
any of the following:
    (i) Any event or condition that results in--
    (A) The condition of the commercial nuclear plant, including its 
principal safety barriers, being seriously degraded; or
    (B) The commercial nuclear plant being in a condition not analyzed 
under Sec.  53.450 that significantly degrades plant safety.
    (ii) Any event or condition that results in valid actuation of an 
SR system, except when the actuation results from and is part of a pre-
planned sequence during testing or reactor operation.
    (iii) Any event or condition that at the time of discovery could 
have prevented the fulfilment of the safety functions identified under 
Sec.  53.230. Events covered may include one or more procedural errors, 
equipment failures, and/or discovery of design, analysis, fabrication, 
construction, and/or procedural inadequacies. However, individual 
component failures need not be reported pursuant to this paragraph if 
other equipment was operable and available to perform the required 
safety function.
    (iv) Any event requiring the transport of a radioactively 
contaminated person to an offsite medical facility for treatment.
    (v) Any event that results in a major loss of emergency assessment 
capability, offsite response capability, or offsite communications 
capability (e.g., significant portion of control room indication, ENS, 
or offsite notification system).
    (c) Follow-up notification: With respect to the notifications made 
under paragraphs (a) and (b) of this section, in addition to making the 
required initial notification, each licensee, must during the course of 
the event--
    (1) Immediately Report:
    (i) any further degradation in the level of safety of the plant or 
other worsening plant conditions, including those that require the 
declaration of any of the Emergency Classes, if such a declaration has 
not been previously made, or
    (ii) any change from one Emergency Class to another, or
    (iii) a termination of the Emergency Class.
    (2) Immediately Report:
    (i) the results of ensuing evaluations or assessments of plant 
conditions,
    (ii) the effectiveness of response or protective measures taken, 
and
    (iii) important information related to plant behavior that is not 
understood.
    (3) Maintain an open, continuous communication channel with the NRC 
Operation Center upon request by the NRC.

    \1\ Other requirements for immediate notification of the NRC by 
licensed operating commercial nuclear plants are contained elsewhere 
in this chapter, in particular Sec. Sec.  20.1906, 20.2202, 72.216, 
73.77, and 73.1200 of this chapter.


Sec.  53.1640  Licensee event report system.

    (a) Reportable events. (1) Each commercial nuclear plant licensee 
holding an OL under this part or a COL under this part after the 
Commission makes the finding under Sec.  53.1452(g), must submit a 
Licensee Event Report (LER) for any event of the type described in this 
paragraph within 60 days after discovery of the event. In the case of 
an invalid actuation reported under Sec.  53.1640(a)(2), other than 
automatic reactor shutdown when the reactor is critical, the licensee 
may, at its option, provide a telephone notification to the NRC 
Operations Center within 60 days after discovery of the event instead 
of submitting a written LER. Unless otherwise specified in this 
section, the licensee must report an event if it occurred within 3 
years of the date of discovery regardless of the plant mode or power 
level, and regardless of the significance of the structure, system, or 
component that initiated the event.
    (2) The licensee must report--
    (i)(A) The completion of any commercial nuclear plant shutdown 
required by the plant's Technical Specifications.

[[Page 87113]]

    (B) Any operation or condition which was prohibited by the plant's 
Technical Specifications except when--
    (1) The Technical Specification is administrative in nature;
    (2) The event consisted solely of a case of a late surveillance 
test where the oversight was corrected, the test was performed, and the 
equipment was found to be capable of performing its specified safety 
functions; or
    (3) The Technical Specification was revised prior to discovery of 
the event such that the operation or condition was no longer prohibited 
at the time of the event.
    (C) Any deviation from the plant's Technical Specifications 
authorized under Sec.  53.740(h).
    (ii) Any event or condition that resulted in--
    (A) The condition of the commercial nuclear plant, including its 
principal safety barriers, being seriously degraded; or
    (B) The commercial nuclear plant being in a condition not analyzed 
under Sec.  53.450 that significantly degrades plant safety.
    (iii) Any natural phenomena or other external condition that posed 
an actual threat to the safety of the commercial nuclear plant or 
significantly hampered site personnel in the performance of duties 
necessary for the safe operation of the commercial nuclear plant.
    (iv) Any event or condition that resulted in inadvertent operation 
of any structures, systems, and component classified as SR for an 
identified safety function under Sec.  53.460 or the unplanned sole 
reliance on an SR system for those systems that are in constant 
operation, except when--
    (A) The actuation resulted from and was part of a pre-planned 
sequence during testing; or
    (B) The actuation was invalid and--
    (1) Occurred while the system was properly removed from service; or
    (2) Occurred after the safety function had been already completed.
    (v) Any event or condition that could have prevented the 
fulfillment of the safety functions identified under Sec.  53.230.
    (vi) Events covered in paragraph (a)(2)(v) of this section may 
include one or more procedural errors, equipment failures, and/or 
discovery of design, fabrication, construction, and/or procedural 
inadequacies. However, individual component failures need not be 
reported pursuant to paragraph (a)(2)(v) of this section if any other 
equipment was operable and available to perform the required safety 
function.
    (vii)(A) Any event or condition that as a result of a single cause 
could have prevented the fulfillment of any of the safety functions 
identified under Sec.  53.230.
    (B) Events covered in paragraph (a)(2)(vii)(A) of this section may 
include cases of procedural error, equipment failure, and/or discovery 
of a design, analysis, fabrication, construction, and/or procedural 
inadequacy. However, licensees are not required to report an event 
pursuant to paragraph (a)(2)(vii)(A) of this section if the event 
results from--
    (1) A shared dependency among trains or channels that is a natural 
or expected consequence of the approved plant design; or
    (2) Normal and expected wear or degradation.
    (viii)(A) Any airborne radioactive release that, when averaged over 
a time period of 1-hour, resulted in airborne radionuclide 
concentrations in an unrestricted area that exceeds 20 times the 
applicable concentration limits specified in appendix B to 10 CFR part 
20, table 2, column 1.
    (B) Any liquid effluent release that, when averaged over a time 
period of 1-hour, exceeds 20 times the applicable concentrations 
specified in appendix B to 10 CFR part 20, table 2, column 2, at the 
point of entry into the receiving waters (i.e., unrestricted area) for 
all radionuclides except tritium and dissolved noble gases.
    (ix) Any event that posed an actual threat to the safety of the 
commercial nuclear plant or significantly hampered site personnel in 
the performance of duties necessary for the safe operation of the 
plant, including fires, toxic gas releases, or radioactive releases.
    (b) Contents. The LER must contain--
    (1) A brief abstract describing the major occurrences during the 
event, including all component or system failures that contributed to 
the event and significant corrective action taken or planned to prevent 
recurrence.
    (2)(i) A clear, specific narrative description of what occurred so 
that knowledgeable readers conversant with the design of commercial 
nuclear plants, but not familiar with the details of a particular 
plant, can understand the complete event.
    (ii) The narrative description must include the following specific 
information as appropriate for the particular event:
    (A) Plant operating conditions before the event.
    (B) Status of systems, structures, or components that were 
inoperable at the start of the event and that contributed to the event.
    (C) Dates and approximate time of the occurrences.
    (D) The cause of each component or system failure or personnel 
error, if known.
    (E) The failure mode, mechanism, and effect of each failed 
component, if known.
    (F) [Reserved]
    (G) For failures of components with multiple functions, include a 
list of systems or secondary functions that were also affected.
    (H) For failure that rendered a component or system classified as 
SR or non-safety-related but safety-significant inoperable, an estimate 
of the elapsed time from the discovery of the failure until the 
component or system was returned to service.
    (I) The method of discovery of each component or system failure or 
procedural error.
    (J) For each human performance related root cause, the licensee 
must discuss the cause(s) and circumstances.
    (K) Automatically and manually initiated safety system responses.
    (L) The manufacturer and model number (or other identification) of 
each component that failed during the event.
    (3) An assessment of the safety consequences and implications of 
the event. This assessment must include--
    (i) The availability of systems or components that could have 
performed the same function as the components and systems that failed 
during the event, and
    (ii) For events that occurred when the reactor was shut down, the 
availability of systems or components that are needed to shut down the 
reactor and maintain safe shutdown conditions, remove residual heat, 
control the release of radioactive material, or mitigate the 
consequences of an accident.
    (4) A description of any corrective actions planned as a result of 
the event, including those to reduce the probability of similar events 
occurring in the future.
    (5) Reference to any previous similar events at the same plant that 
are known to the licensee.
    (6) The name and contact information of a person within the 
licensee's organization who is knowledgeable about the event and can 
provide additional information concerning the event and the plant's 
characteristics.
    (c) Supplemental Information. The Commission may require the 
licensee to submit specific additional information beyond that required 
by paragraph (b) of this section if the Commission finds that 
supplemental material is necessary for complete understanding of an 
unusually complex or significant event. These requests for supplemental 
information will be made in writing and the licensee

[[Page 87114]]

must submit, as specified in Sec.  53.040, the requested information as 
a supplement to the initial LER.
    (d) Submission of Reports. Licensee Event Reports must be prepared 
on Form NRC 366 and submitted to the NRC, as specified in Sec.  53.040.
    (e) Report Legibility. The reports and copies that licensees are 
required to submit to the Commission under the provisions of this 
section must be of sufficient quality to permit legible reproduction 
and micrographic processing.


53.1645  Reports of radiation exposure to members of the public.

    (a) Each holder of an OL, and each holder of a COL after the 
Commission has made the finding under Sec.  53.1452(g), must submit 
radiological reports as required by 10 CFR part 20, as well as an 
Annual Radioactive Effluent Release Report and an Annual Radiological 
Environmental Operating Report. The Annual Radioactive Effluent Release 
Report must specify the quantity of each of the principal radionuclides 
released to unrestricted areas in liquid and in gaseous effluents and 
an estimate of the dose received by the maximally exposed member of the 
public in an unrestricted area from effluents and direct radiation from 
contained sources during the previous calendar year. The Annual 
Radiological Environmental Operating Report must provide data on 
measurable levels of radiation and radioactive materials in the 
environment, must include an evaluation of the relationship between 
quantities of radioactive material released in effluents and resultant 
radiation doses to individuals from principal pathways of exposure, and 
must include the results of environmental monitoring during the 
previous calendar year. These reports must also include any other 
information as may be required by the Commission to estimate maximum 
potential annual radiation doses to the public. The reports must be 
submitted as specified in Sec.  53.040 by May 15 of each successive 
year. If the total effective dose equivalent to members of the public 
in unrestricted areas during the reporting period is greater than the 
as low as is reasonably achievable (ALARA) design objectives 
established under Sec.  53.425, the report must specify the causes for 
exceeding the ALARA design objective and describe any corrective 
actions. On the basis of these reports and any additional information 
the Commission may obtain from the licensee or others, the Commission 
may require the licensee to take action as the Commission deems 
appropriate.
    (b) If during any calendar quarter the radiation exposure to a 
member of the public in the unrestricted areas, calculated on the same 
basis as the respective ALARA design objective exposure, exceeds one-
half of the annual ALARA design objective exposure, the licensee must 
submit a report as specified in Sec.  53.040. The report shall specify 
the causes for exceeding one-half the annual ALARA design objective 
exposure in a quarter and describe corrective actions that the licensee 
will take to maintain radiation exposure to levels within the ALARA 
design objectives for the remainder of the year. The report shall be 
submitted within 30 days from the end of the quarter when one-half of 
the annual ALARA design objective exposure was exceeded.


Sec.  53.1650  Facility information and verification.

    (a) In response to a written request by the Commission, each 
applicant for a CP or license and each recipient of a CP or a license 
must submit facility information, as described in Sec.  75.10 of this 
chapter, on International Atomic Energy Agency (IAEA) Design 
Information Questionnaire forms and site information on DOC/NRC Form 
AP-A and associated forms;
    (b) As required by the Additional Protocol, must submit location 
information described in Sec.  75.11 of this chapter on DOC/NRC Form 
AP-1 and associated forms; and
    (c) Must permit verification thereof by the IAEA and take other 
action as necessary to implement the US/IAEA Safeguards Agreement, as 
described in part 75 of this chapter.


Sec.  53.1660  Financial requirements.

    Sections 53.1670 through 53.1700 set out the requirements and 
procedures related to financial qualifications and related reporting 
requirements.


Sec.  53.1670  Financial qualifications.

    Except for an electric utility applicant for a license to operate a 
commercial nuclear plant, an applicant for a CP, OL, or COL under this 
part must possess or have reasonable assurance of obtaining the funds 
necessary for the activities for which the permit or license is sought.


Sec.  53.1680  Annual financial reports.

    With respect to any commercial nuclear plant of a type described in 
Sec.  53.020, each licensee and each holder of a CP must submit its 
annual financial report, including the certified financial statements, 
to the Commission, as specified in Sec.  53.040, upon issuance of the 
report. However, licensees and holders of a CP who submit a Form 10-Q 
with the Securities and Exchange Commission or a Form 1 with the 
Federal Energy Regulatory Commission need not submit the annual 
financial report or the certified financial statement under this 
section.


Sec.  53.1690  Licensee's change of status; financial qualifications.

    (a) An electric utility licensee holding an OL or COL (including a 
renewed license) for a commercial nuclear plant, no later than seventy-
five (75) days prior to ceasing to be an electric utility in any manner 
not involving a license transfer under Sec.  53.1399 or Sec.  53.1456 
must provide the NRC with the financial qualifications information that 
would be required for obtaining an initial OL or COL under this part. 
The financial qualifications information must address the first full 5 
years of operation after the date the licensee ceases to be an electric 
utility.
    (b)(1) Any holder of a license issued under this part must notify 
the appropriate NRC Regional Administrator, in writing, immediately 
following the filing of a voluntary or involuntary petition for 
bankruptcy under any chapter of title 11 (Bankruptcy) of the United 
States Code by or against--
    (i) The licensee;
    (ii) An entity (as 11 U.S.C. 101(14) defines that term) controlling 
the licensee or listing the license or licensee as property of the 
estate; or
    (iii) An affiliate (as 11 U.S.C. 101(2) defines that term) of the 
licensee.
    (2) This notification must indicate--
    (i) The bankruptcy court in which the petition for bankruptcy was 
filed; and
    (ii) The date of the filing of the petition.


Sec.  53.1700  Creditor regulations.

    (a) Pursuant to section 184 of the Act, the Commission consents, 
without individual application, to the creation of any mortgage, 
pledge, or other lien upon any facility not owned by the United States 
which is the subject of a license or upon any leasehold or other 
interest in such facility; provided--
    (1) That the rights of any creditor so secured may be exercised 
only in compliance with and subject to the same requirements and 
restrictions as would apply to the licensee pursuant to the provisions 
of the license, the Act, and regulations issued by the Commission under 
the Act; and
    (2) That no creditor so secured may take possession of the facility 
pursuant to the provisions of this section prior to either the issuance 
of a license from the Commission authorizing such possession or the 
transfer of the license.

[[Page 87115]]

    (b) Any creditor so secured may apply for transfer of the license 
covering such facility by filing an application for transfer of the 
license under Sec.  53.1570. The Commission will act upon such 
application under subpart I of this part.
    (c) Nothing contained in this regulation shall be deemed to affect 
the means of acquiring, or the priority of, any tax lien or other lien 
provided by law.
    (d) As used in this section--
    License includes any license under this part, which may be issued 
by the Commission with regard to a facility.
    Creditor includes, without implied limitation, the trustee under 
any mortgage, pledge or lien on a facility made to secure any creditor, 
any trustee or receiver of the facility appointed by a court of 
competent jurisdiction in any action brought for the benefit of any 
creditor secured by such mortgage, pledge or lien, any purchaser of 
such facility at the sale thereof upon foreclosure of such mortgage, 
pledge, or lien or upon exercise of any power of sale contained 
therein, or any assignee of any such purchaser.
    Facility includes, but is not limited to, a site which is the 
subject of an early site permit under this part, and a reactor 
manufactured under an ML under this part.


Sec.  53.1710  Financial protection.

    Sections 53.1720 and 53.1730 set out the requirements and 
procedures related to licensees obtaining and maintaining insurance to 
cover stabilization and decontamination activities in the event of an 
accident and financial protection in accordance with part 140, 
``Financial Protection Requirements and Indemnity Agreements,'' of this 
chapter.


Sec.  53.1720  Insurance required to stabilize and decontaminate plant 
following an accident.

    Each commercial nuclear plant licensee under this part must take 
reasonable steps to obtain insurance available at reasonable costs and 
on reasonable terms from private sources or to demonstrate that it 
possesses an equivalent amount of protection covering the licensee's 
obligation, in the event of an accident at the licensee's commercial 
nuclear reactor, to stabilize and decontaminate the plant and the plant 
site at which such an accident may occur, provided that--
    (a) The insurance required by this section must have a minimum 
coverage limit for each commercial nuclear plant site of $1.06 billion, 
an amount based on plant-specific estimates of costs to stabilize and 
decontaminate a plant, or whatever amount of insurance is generally 
available from private sources, whichever is less. The required 
insurance must clearly state that, as and to the extent provided in 
paragraph (d)(1) of this section, any proceeds must be payable first 
for stabilization of the plant and next for decontamination of the 
plant and the plant site. If a licensee's coverage falls below the 
required minimum, the licensee must within 60 days take all reasonable 
steps to restore its coverage to the required minimum. The required 
insurance may, at the option of the licensee, be included within 
policies that also provide coverage for other risks, including, but not 
limited to, the risk of direct physical damage.
    (b)(1) With respect to policies issued or annually renewed, the 
proceeds of such required insurance must be dedicated, as and to the 
extent provided in this paragraph, to reimbursement or payment on 
behalf of the insured of reasonable expenses incurred or estimated to 
be incurred by the licensee in taking action to fulfill the licensee's 
obligation, in the event of an accident at the licensee's plant, to 
ensure that the plant is in, or is returned to, and maintained in, a 
safe and stable condition and that radioactive contamination is removed 
or controlled such that personnel exposures are consistent with the 
occupational exposure limits in 10 CFR part 20. These actions must be 
consistent with any other obligation the licensee may have under this 
chapter and must be subject to paragraph (d) of this section. As used 
in this section, an ``accident'' means an event that involves the 
release of radioactive material from its intended place of confinement 
within the commercial nuclear plant such that there is a present danger 
of release off site in amounts that would pose a threat to the public 
health and safety.
    (2) The stabilization and decontamination requirements set forth in 
paragraph (d) of this section must apply uniformly to all insurance 
policies required under this section.
    (c) The licensee shall report to the NRC on April 1 of each year 
the current levels of this insurance or financial security it maintains 
and the sources of this insurance or financial security.
    (d)(1) In the event of an accident at the licensee's plant, 
whenever the estimated costs of stabilizing the licensed plant and of 
decontaminating the plant and the plant site exceed one tenth of the 
minimum insurance under paragraph (a) of this section, the proceeds of 
the insurance required by this section must be dedicated to and used, 
first, to ensure that the licensed plant is in, or is returned to, and 
can be maintained in, a safe and stable condition so as to prevent any 
significant risk to the public health and safety and, second, to 
decontaminate the plant and the plant site in accordance with the 
licensee's cleanup plan as approved by order of the Director, Office of 
Nuclear Reactor Regulation. This priority on insurance proceeds must 
remain in effect for 60 days or, upon order of the Director, for such 
longer periods, in increments not to exceed 60 days except as provided 
for activities under the cleanup plan required in paragraphs (d)(3) and 
(d)(4) of this section, as the Director may find necessary to protect 
the public health and safety. Actions needed to bring the plant to and 
maintain the plant in a safe and stable condition may include one or 
more of the following, as appropriate:
    (i) Shutdown of the reactor(s) and other processes at the plant;
    (ii) Establishment and maintenance of long-term cooling with stable 
decay heat removal;
    (iii) Maintenance of sub-criticality;
    (iv) Control of radioactive releases; and
    (v) Securing of structures, systems, or components to minimize 
radiation exposure to onsite personnel or to the offsite public or to 
facilitate later decontamination or both.
    (2) The licensee must inform the Director, Office of Nuclear 
Reactor Regulation in writing when the plant is and can be maintained 
in a safe and stable condition so as to prevent any significant risk to 
the public health and safety. Within 30 days after the licensee informs 
the Director that the plant is in this condition, or at such earlier 
time as the licensee may elect or the Director may for good cause 
direct, the licensee must prepare and submit a cleanup plan for the 
Director's approval. The cleanup plan must identify and contain an 
estimate of the cost of each cleanup operation that will be required to 
decontaminate the reactor sufficiently to permit the licensee either to 
resume operation of the reactor or to apply to the Commission under 
subpart G of this part for authority to decommission the reactor and to 
surrender the license voluntarily. Cleanup operations may include one 
or more of the following, as appropriate:
    (i) Processing any contaminated materials generated by the accident 
and by decontamination operations to remove radioactive materials;
    (ii) Decontamination of surfaces inside the plant buildings to 
levels consistent with the Commission's occupational exposure limits in 
10 CFR part 20, and decontamination or disposal of equipment;

[[Page 87116]]

    (iii) Decontamination or removal and disposal of internal parts, 
damaged fuel from the reactor coolant or fuel systems, or related 
process or waste systems; and
    (iv) Cleanup of the reactor coolant or fuel systems or related 
process or waste systems.
    (3) Following review of the licensee's cleanup plan, the Director 
will order the licensee to complete all operations that the Director 
finds are necessary to decontaminate the reactor sufficiently to permit 
the licensee either to resume operation of the reactor or to apply to 
the Commission under subpart G of this part for authority to 
decommission the reactor and to surrender the license voluntarily. The 
Director must approve or disapprove, in whole or in part for stated 
reasons, the licensee's estimate of cleanup costs for such operations. 
Such order may not be effective for more than one year, at which time 
it may be renewed. Each subsequent renewal order, if imposed, may be 
effective for not more than 6 months.
    (4) Of the balance of the proceeds of the required insurance not 
already expended to place the plant in a safe and stable condition 
under paragraph (b)(1) of this section, an amount sufficient to cover 
the expenses of completion of those decontamination operations that are 
the subject of the Director's order must be dedicated to such use, 
provided that, upon certification to the Director of the amounts 
expended previously and from time to time for stabilization and 
decontamination and upon further certification to the Director as to 
the sufficiency of the dedicated amount remaining, policies of 
insurance may provide for payment to the licensee or other loss payees 
of amounts not so dedicated, and the licensee may proceed to use in 
parallel (and not in preference thereto) any insurance proceeds not so 
dedicated for other purposes.


Sec.  53.1730  Financial protection requirements.

    Commercial nuclear plant licensees must satisfy the applicable 
provisions of part 140, ``Financial Protection Requirements and 
Indemnity Agreements,'' of this chapter.

Subparts K and L [Reserved]

Subpart M--Enforcement


Sec.  53.9000  Violations.

    (a) The Commission may obtain an injunction or other court order to 
prevent a violation of the provisions of--
    (1) The Atomic Energy Act of 1954, as amended (the Act);
    (2) Title II of the Energy Reorganization Act of 1974, as amended; 
or
    (3) A regulation or order issued under those Acts.
    (b) The Commission may obtain a court order for the payment of a 
civil penalty imposed under Section 234 of the Act:
    (1) For violations of--
    (i) Sections 53, 57, 62, 63, 81, 82, 101, 103, 104, 107, or 109 of 
the Act;
    (ii) Section 206 of the Energy Reorganization Act of 1974, as 
amended;
    (iii) Any rule, regulation, or order issued under the sections 
specified in paragraph (b)(1)(i) of this section;
    (iv) Any term, condition, or limitation of any license issued under 
the sections specified in paragraph (b)(1)(i) of this section.
    (2) For any violation for which a license may be revoked under 
section 186 of the Act.


Sec.  53.9010  Criminal penalties.

    (a) Section 223 of the Act provides for criminal sanctions for 
willful violation of, attempted violation of, or conspiracy to violate, 
any regulation issued under sections 161b, 161i, or 161o of the Act. 
For purposes of section 223, all the regulations in part 53 are issued 
under one or more of sections 161b, 161i, or 161o, except for the 
sections listed in paragraph (b) of this section.
    (b) The regulations in 10 CFR part 53 that are not issued under 
sections 161b, 161i, or 161o for the purposes of section 223 are as 
follows: Sec. Sec.  53.000, 53.015, 53.020, 53.040, 53.080, 53.090, 
53.100, 53.110, 53.120, 53.600, 53.725, 53.726, 53.735, 53.760, 53.775, 
53.790, 53.795, 53.820, 53.910, 53.1000, 53.1050, 53.1100, 53.1103, 
53.1106, 53.1109, 53.1112, 53.1115, 53.1118, 53.1120, 53.1121, 53.1124, 
53.1140, 53.1143, 53.1144, 53.1146, 53.1149, 53.1155, 53.1158, 53.1164, 
53.1170, 53.1173, 53.1176, 53.1179, 53.1188, 53.1200, 53.1203, 53.1206, 
53.1209, 53.1210, 53.1212, 53.1215, 53.1218, 53.1221, 53.1230, 53.1236, 
53.1239, 53.1241, 53.1242, 53.1245, 53.1248, 53.1251, 53.1254, 52.1257, 
52.1260, 53.1263, 53.1270, 53.1273, 53.1276, 53.1279, 53.1282, 53.1285, 
53.1286, 53.1287, 53.1288, 53.1291, 53.1293, 53.125, 53.1300, 53.1306, 
53.1309, 53.1312, 53.1315, 53.1318, 53.1324, 53.1330, 53.1333, 53.1336, 
53.1348, 53.1360, 53.1366, 53.1369, 53.1372, 53.1375, 53.1381, 53.1384, 
53.1387, 53.1390, 53.1396, 53.1401, 53.1405, 53.1410, 53.1416, 53.1419, 
53.1422, 53.1425, 53.1431, 53.1437, 53.1440, 53.1443, 53.1452, 53.1455, 
53.1456, 53.1458, 53.1461, 53.1470, 53.1500, 53.1510, 53.1515, 53.1520, 
53.1525, 53.1530, 53.1535, 53.1540, 53.1560, 53.1585, 53.1590, 53.1595, 
53.1600, 53.1660, 53.1670, 53.1700, 53.1710, 53.1730, 53.9000, 53.9010.

PART 70--DOMESTIC LICENSING OF SPECIAL NUCLEAR MATERIAL

0
129. The authority citation for 10 CFR part 70 continues to read as 
follows:

    Authority:  Atomic Energy Act of 1954, secs. 51, 53, 57(d), 108, 
122, 161, 182, 183, 184, 186, 187, 193, 223, 234, 274, 1701 (42 
U.S.C. 2071, 2073, 2077(d), 2138, 2152, 2201, 2232, 2233, 2234, 
2236, 2237, 2243, 2273, 2282, 2021, 2297f); Energy Reorganization 
Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 
5851); Nuclear Waste Policy Act of 1982, secs. 135, 141 (42 U.S.C. 
10155, 10161); 44 U.S.C. 3504 note.


Sec.  70.20a  [Amended]

0
130. In Sec.  70.20a, in paragraph (b) remove the phrase ``parts 30 
through 36, 39, 40, 50, 72, 110,'' and add in its place the phrase 
``parts 30 through 36, 39, 40, 50, 53, 72, 110''.


Sec.  70.22  [Amended]

0
131. In Sec.  70.22, wherever it appears, remove the phrase ``part 50'' 
and add in its place the phrase ``parst 50 or 53''.
0
132. In Sec.  70.24, revise paragraph (d) to read as follows:


Sec.  70.24  Criticality accident requirements.

* * * * *
    (d)(1) The requirements in paragraphs (a) through (c) of this 
section do not apply to a holder of a construction permit or operating 
license for a nuclear power reactor issued under part 50 or part 53 of 
this chapter or a combined license issued under part 52 or part 53 of 
this chapter, if the holder complies with the requirements of paragraph 
(b) of 10 CFR 50.68 or paragraph (m)(2) of 10 CFR 53.440,as applicable.
    (2) An exemption from Sec.  70.24 held by a licensee who thereafter 
elects to comply with requirements of paragraph (b) of 10 CFR 50.68 or 
paragraph (m)(2) of 10 CFR 53.440 does not exempt that licensee from 
complying with any of the requirements in Sec.  50.68 or Sec.  
53.440(m) of this chapter but shall be ineffective so long as the 
licensee elects to comply with Sec.  50.68(b) or Sec.  53.440(m)(2) of 
this chapter, as applicable.


Sec.  70.32  [Amended]

0
133. In Sec.  70.32, in paragraph (c)(1) introductory text, remove the 
phrase ``part 50 of this chapter'' and add in its place the phrase 
``parts 50 or 53 of this chapter''; and in paragraph (d) remove the 
phrase ``or Sec.  70.34 of this chapter, as appropriate.'' and add in 
its place the phrase ``, Sec. Sec.  74.34 or 53.1510 of this chapter, 
as appropriate.''.

[[Page 87117]]

0
134. In Sec.  70.50, revise paragraph (d) to read as follows:


Sec.  70.50  Reporting requirements.

* * * * *
    (d) The provisions of Sec.  70.50 do not apply to licensees subject 
to Sec. Sec.  50.72 or 53.1630 of this chapter. They do apply to those 
10 CFR parts 50 or 53 licensees possessing material licensed under 10 
CFR part 70 that are not subject to the notification requirements in 
Sec. Sec.  50.72 or 53.1630 of this chapter.

PART 72--LICENSING REQUIREMENTS FOR THE INDEPENDENT STORAGE OF 
SPENT NUCLEAR FUEL AND HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-
RELATED GREATER THAN CLASS C WASTE

0
135. The authority citation for 10 CFR part 72 continues to read as 
follows:

    Authority:  Atomic Energy Act of 1954, secs. 51, 53, 57, 62, 63, 
65, 69, 81, 161, 182, 183, 184, 186, 187, 189, 223, 234, 274 (42 
U.S.C. 2071, 2073, 2077, 2092, 2093, 2095, 2099, 2111, 2201, 2210e, 
2232, 2233, 2234, 2236, 2237, 2238, 2273, 2282, 2021); Energy 
Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C. 
5841, 5842, 5846, 5851); National Environmental Policy Act of 1969 
(42 U.S.C. 4332); Nuclear Waste Policy Act of 1982, secs. 117(a), 
132, 133, 134, 135, 137, 141, 145(g), 148, 218(a) (42 U.S.C. 
10137(a), 10152, 10153, 10154, 10155, 10157, 10161, 10165(g), 10168, 
10198(a)); 44 U.S.C. 3504 note.

0
136. In Sec.  72.3, revise the definition for ``Independent spent fuel 
storage installation or ISFSI'' to read as follows:


Sec.  72.3  Definitions.

* * * * *
    Independent spent fuel storage installation or ISFSI means a 
complex designed and constructed for the interim storage of spent 
nuclear fuel, solid reactor-related GTCC waste, and other radioactive 
materials associated with spent fuel and reactor-related GTCC waste 
storage. An ISFSI which is located on the site of another facility 
licensed under this part or a facility licensed under part 50 or part 
53 of this chapter and which shares common utilities and services with 
that facility or is physically connected with that other facility may 
still be considered independent.
* * * * *
0
137. In Sec.  72.30, revise paragraph (e)(5) to read as follows:


Sec.  72.30  Financial assurance and recordkeeping for decommissioning.

* * * * *
    (e) * * *
    (5) In the case of licensees who are issued a power reactor license 
under parts 50 or 53 of this chapter or ISFSI licensees who are an 
electric utility, as defined in parts 50 or 53 of this chapter, with a 
specific license issued under this part, the methods of Sec. Sec.  
50.75(b), (e), and (h) or 53.1010, 53.1040, 53.1045(b), and 53.1060 of 
this chapter, as applicable. In the event that funds remaining to be 
placed into the licensee's ISFSI decommissioning external sinking fund 
are no longer approved for recovery in rates by a competent rate making 
authority, the licensee must make changes to provide financial 
assurance using one or more of the methods stated in paragraphs (a)(1) 
through (4) of this section.
* * * * *
0
138. In Sec.  72.32, revise paragraph (c)(2) to read as follows:


Sec.  72.32  Emergency plan.

* * * * *
    (c) * * *
    (2)(i) Located within the exclusion area as defined in 10 CFR part 
100, of a nuclear power reactor licensed for operation by the 
Commission, the emergency plan that meets either the requirements in 
Sec.  50.160 of this chapter or the requirements in appendix E to part 
50 of this chapter and Sec.  50.47(b) of this chapter shall be deemed 
to satisfy the requirements of this section.
    (ii) Located within the exclusion area, as defined in 10 CFR part 
53, of a commercial nuclear plant licensed for operation by the 
Commission, the emergency plan that meets either the requirements in 
Sec.  50.160 of this chapter or the requirements in appendix E to part 
50 of this chapter and Sec.  50.47(b) of this chapter shall be deemed 
to satisfy the requirements of this section.
* * * * *


Sec.  72.40  [Amended]

0
139. In Sec.  72.40, in paragraph (c) remove the phrase ``under part 50 
of this chapter,'' and add in its place the phrase ``under parts 50 or 
53 of this chapter,''.
0
140. In Sec.  72.75, revise paragraph (i)(1)(ii) to read as follows:


Sec.  72.75  Reporting requirements for specific events and conditions.

* * * * *
    (i) * * *
    (1) * * *
    (ii) Licensees issued a general license under Sec.  72.210, after 
the licensee has placed spent fuel on the ISFSI storage pad (if the 
ISFSI is located inside the collocated protected area, for a reactor 
licensed under parts 50 or 53 of this chapter) or after the licensee 
has transferred spent fuel waste outside the reactor licensee's 
protected area to the ISFSI storage pad (if the ISFSI is located 
outside the collocated protected area, for a reactor licensed under 
parts 50 or 53 of this chapter).
* * * * *


Sec.  72.184  [Amended]

0
141. In Sec.  72.184, in paragraph (a) remove the phrase ``under part 
50 of this chapter'' and add in its place the phrase ``under parts 50 
or 53 of this chapter''.
0
142. Revise Sec.  72.210 to read as follows:


Sec.  72.210  General license issued.

    A general license is hereby issued for the storage of spent fuel in 
an independent spent fuel storage installation at power reactor sites 
to persons authorized to possess or operate nuclear power reactors 
under 10 CFR parts 50, 52, or 53.
0
143. In Sec.  72.212, revise paragraph (b)(8) to read as follows:


Sec.  72.212  Conditions of general license issued under Sec.  72.210.

* * * * *
    (b) * * *
    (8) Before use of the general license, determine whether activities 
related to storage of spent fuel under this general license involve a 
change in the facility Technical Specifications or require a license 
amendment for the facility pursuant to Sec. Sec.  50.59(c) or 53.1550 
of this chapter. Results of this determination must be documented in 
the evaluations made in paragraph (b)(5) of this section.
* * * * *
0
144. In Sec.  72.218, revise paragraphs (a) and (b) to read as follows:


Sec.  72.218  Termination of licenses.

    (a) The notification regarding the program for the management of 
spent fuel at the reactor required by Sec. Sec.  50.54(bb) or 53.1060 
of this chapter must include a plan for removal of the spent fuel 
stored under this general license from the reactor site. The plan must 
show how the spent fuel will be managed before starting to decommission 
systems and components needed for moving, unloading, and shipping this 
spent fuel.
    (b) An application for termination of a reactor operating license 
issued under 10 CFR part 50 and submitted under Sec.  50.82 of this 
chapter, or a combined license issued under 10 CFR part 52 and 
submitted under Sec.  52.110 of this chapter, or a reactor operating or 
combined license under 10 CFR part 53 and submitted under Sec.  53.1070 
of this chapter must contain a description of how the spent fuel stored 
under this

[[Page 87118]]

general license will be removed from the reactor site.
* * * * *

PART 73--PHYSICAL PROTECTION OF PLANTS AND MATERIALS

0
145. The authority citation for 10 CFR part 73 continues to read as 
follows:

    Authority:  Atomic Energy Act of 1954, secs. 53, 147, 149, 161, 
161A, 170D, 170E, 170H, 170I, 223, 229, 234, 1701 (42 U.S.C. 2073, 
2167, 2169, 2201, 2201a, 2210d, 2210e, 2210h, 2210i, 2273, 2278a, 
2282, 2297f); Energy Reorganization Act of 1974, secs. 201, 202 (42 
U.S.C. 5841, 5842); Nuclear Waste Policy Act of 1982, secs. 135, 141 
(42 U.S.C. 10155, 10161); 44 U.S.C. 3504 note.

    Section 73.37(b)(2) also issued under sec. 301, Pub. L. 96-295, 
94 Stat. 789 (42 U.S.C. 5841 note).

0
146. In Sec.  73.1, revise paragraph (b)(1)(i) to read as follows:


Sec.  73.1  [Amended]

* * * * *
    (b) * * *
    (1) * * *
    (i) The physical protection of production and utilization 
facilities licensed under parts 50, 52, or 53 of this chapter,
* * * * *
0
147. In Sec.  73.2, revise the introductory text and paragraph (a) to 
read as follows:


Sec.  73.2  Definitions.

    As used in this part:
    (a) Terms defined in parts 50, 52, 53, 70, and 95 of this chapter 
have the same meaning when used in this part.
* * * * *
0
148. In Sec.  73.8, revise paragraph (b) to read as follows:


Sec.  73.8  Information collection requirements: OMB approval.

* * * * *
    (b) The approved information collection requirements contained in 
this part appear in Sec. Sec.  73.5, 73.15, 73.17, 73.20, 73.21, 73.24, 
73.25, 73.26, 73.27, 73.37, 73.40, 73.45, 73.46, 73.50, 73.54, 73.55, 
73.56, 73.57, 73.58, 73.60, 73.67, 73.70, 73.72, 73.73, 73.74, 73.77, 
73.100, 73.110, 73.120, 73.1200, 73.1205, 73.1210, 73.1215, and 
appendices B and C to this part.
* * * * *
0
149. In Sec.  73.50, revise the introductory text to read as follows:


Sec.  73.50  Requirements for physical protection of licensed 
activities.

    Each licensee who is not subject to Sec.  73.51, but who possesses, 
uses, or stores formula quantities of strategic special nuclear 
material that are not readily separable from other radioactive material 
and which have a total external radiation level in excess of 1 gray 
(100 rad) per hour at a distance of 1 meter (3.3 feet) from any 
accessible surfaces without intervening shielding other than at a 
nuclear reactor facility licensed under parts 50, 52, or 53 of this 
chapter, shall comply with the following:
* * * * *
0
150. In Sec.  73.55, revise paragraphs (a)(4) and (6), (i)(4)(iii), 
(l)(1), (l)(7)(ii), (p)(1)(i), (r)(2) and (r)(4)(iii) to read as 
follows:


Sec.  73.55  Requirements for physical protection of licensed 
activities in nuclear power reactors against radiological sabotage.

    (a) * * *
    (4) Applicants for an operating license under the provisions of 
part 50 or part 53 of this chapter or holders of a combined license 
under the provisions of part 52 or part 53 of this chapter shall 
implement the requirements of this section before fuel is allowed 
onsite (protected area).
* * * * *
    (6) Applicants for an operating license under the provisions of 
part 50 or part 53 of this chapter, or holders of a combined license 
under the provisions of part 52 or part 53 of this chapter that do not 
reference a standard design certification or reference a standard 
design certification issued after May 26, 2009, shall meet the 
requirement of Sec.  73.55(i)(4)(iii).
* * * * *
    (i) * * *
    (4) * * *
    (iii) Applicants for an operating license under the provisions of 
part 50 of this chapter, or holders of a combined license under the 
provisions of part 52 of this chapter, or licensees under part 53 of 
this chapter that elect to demonstrate compliance with Sec.  73.55, 
consistent with Sec.  53.860(a)(2) of this chapter, shall construct, 
locate, protect, and equip both the central and secondary alarm 
stations to the standards for the central alarm station contained in 
this section. Both alarm stations shall be equal and redundant, such 
that all functions needed to satisfy the requirements of this section 
can be performed in both alarm stations.
* * * * *
    (l) * * *
    (1) Commercial nuclear power reactors licensed under 10 CFR parts 
50, 52, or 53 and authorized to use special nuclear material in the 
form of MOX fuel assemblies containing up to 20 weight percent 
PuO2 shall, in addition to demonstrating compliance with the 
requirements of this section, protect un-irradiated MOX fuel assemblies 
against theft or diversion as described in this paragraph.
* * * * *
    (7) * * *
    (ii) Additional measures for the physical protection of un-
irradiated MOX fuel assemblies containing greater than 20 weight 
percent PuO2 shall be determined by the Commission on a 
case-by-case basis and documented through license amendment in 
accordance with Sec. Sec.  50.90 or 53.1510 of this chapter.
* * * * *
    (p) * * *
    (1) * * *
    (i) Under Sec. Sec.  50.54(x) and (y) or 53.740(h) of this chapter, 
the licensee may suspend any security measures under this section in an 
emergency when this action is immediately needed to protect the public 
health and safety and no action consistent with license conditions and 
technical specifications that can provide adequate or equivalent 
protection is immediately apparent. This suspension of security 
measures must be approved as a minimum by a licensed senior operator 
before taking this action.
* * * * *
    (r) * * *
    (2) The licensee shall submit proposed alternative measure(s) to 
the Commission for review and approval under Sec. Sec.  50.4 and 50.90, 
or Sec. Sec.  53.040 and 53.1510 of this chapter before implementation.
* * * * *
    (4) * * *
    (iii) Based on comparison of the costs of the alternative measures 
to the costs of demonstrating compliance with the Commission's 
requirements using the essential elements of Sec. Sec.  50.109 or 
53.1590 of this chapter, the costs of fully demonstrating compliance 
with the Commission's requirements are not justified by the protection 
that would be provided.
0
151. In Sec.  73.56, revise paragraph (a)(3) to read as follows:


Sec.  73.56  Personnel access authorization requirements for nuclear 
power plants.

    (a) * * *
    (3) Each applicant for an operating license under the provisions of 
part 50 of this chapter, each holder of a combined license under the 
provisions of part 52 of this chapter, and applicants for an operating 
license or holders of a combined license under part 53 of this chapter 
that do not meet the requirements of Sec.  53.860(a)(2) of this 
chapter, shall implement the

[[Page 87119]]

requirements of this section before fuel is allowed on site (protected 
area).
* * * * *
0
152. In Sec.  73.57, revise paragraph (a)(3) to read as follows:


Sec.  73.57  Requirements for criminal history records checks of 
individuals granted unescorted access to a nuclear power facility, a 
non-power reactor, or access to Safeguards Information.

    (a) * * *
    (3) Before receiving its operating license under 10 CFR parts 50 or 
53 or before the Commission makes its finding under Sec. Sec.  
52.103(g) or 53.1452(g) of this chapter, each applicant for a license 
to operate a nuclear power reactor (including an applicant for a 
combined license) or a non-power reactor may submit fingerprints for 
those individuals who will require unescorted access to the nuclear 
power facility or non-power reactor facility.
* * * * *
0
153. In Sec.  73.58, revise paragraph (a) to read as follows:


Sec.  73.58  Safety/security interface requirements for nuclear power 
reactors.

    (a) Each operating nuclear power reactor licensee with a license 
issued under parts 50, 52, or 53 of this chapter shall comply with the 
requirements of this section.
* * * * *
0
154. In Sec.  73.67, revise paragraphs (d) introductory text and (f) 
introductory text to read as follows:


Sec.  73.67  Licensee fixed site and in-transit requirements for the 
physical protection of special nuclear material of moderate and low 
strategic significance.

* * * * *
    (d) Fixed site requirements for special nuclear material of 
moderate strategic significance. Each licensee who possesses, stores, 
or uses quantities and types of special nuclear material of moderate 
strategic significance at a fixed site or contiguous sites, except as 
allowed by paragraph (b)(2) of this section and except those who are 
licensed to operate a nuclear power reactor pursuant to part 50 or part 
53, provided that the special nuclear material is located within a 
protected area and protected under Sec.  73.55 or Sec.  73.100, shall:
* * * * *
    (f) Fixed site requirements for special nuclear material of low 
strategic significance. Each licensee who possesses, stores, or uses 
special nuclear material of low strategic significance at a fixed site 
or contiguous sites, except those who are licensed to operate a nuclear 
power reactor pursuant to part 50 or part 53, provided that the special 
nuclear material is located within a protected area and protected under 
Sec.  73.55 or Sec.  73.100, shall:
* * * * *
0
155. In Sec.  73.77, revise paragraphs (a), (b)(1), (c)(6) and (7) to 
read as follows:


Sec.  73.77  Cybersecurity event notifications.

    (a) Each licensee subject to the provisions of Sec. Sec.  73.54 or 
73.110 shall notify the NRC Headquarters Operations Center via the 
Emergency Notification System (ENS), under paragraph (c) of this 
section:
    (1) Within one hour after discovery of a cyberattack that adversely 
impacted:
    (i) Safety-related or important-to-safety functions, security 
functions, or emergency preparedness functions (including offsite 
communications); or that compromised support systems and equipment 
resulting in adverse impacts to safety, security, or emergency 
preparedness functions within the scope of Sec.  73.54; or,
    (ii) Functions performed by digital assets that would prevent a 
postulated fission product release resulting in offsite doses exceeding 
the values in Sec.  53.210 of this chapter, or functions performed by 
digital assets used by the licensee for implementing the physical 
security requirements in Sec.  53.860(a) of this chapter.
    (2) Within 4 hours:
    (i) After discovery of a cyberattack that could have caused an 
adverse impact to:
    (A) Safety-related or important-to-safety functions, security 
functions, or emergency preparedness functions (including offsite 
communications); or that could have compromised support systems and 
equipment, which if compromised, could have adversely impacted safety, 
security, or emergency preparedness functions within the scope of Sec.  
73.54; or,
    (B) Functions performed by digital assets that would prevent a 
postulated fission product release resulting in offsite doses exceeding 
the values in Sec.  53.210 of this chapter, or functions performed by 
digital assets used by the licensee for implementing the physical 
security requirements in Sec.  53.860(a) of this chapter.
    (ii) After discovery of a suspected or actual cyberattack initiated 
by personnel with physical or electronic access to digital computer and 
communication systems and networks within the scope of Sec. Sec.  73.54 
or 73.110.
    (iii) After notification of a local, State, or other Federal agency 
(e.g., law enforcement, Federal Bureau of Investigation (FBI), etc.) of 
an event related to the licensee's implementation of their 
cybersecurity program for digital computer and communication systems 
and networks within the scope of Sec. Sec.  73.54 or 73.110 that does 
not otherwise require a notification under paragraph (a) of this 
section.
    (3) Within 8 hours after receipt or collection of information 
regarding observed behavior, activities, or statements that may 
indicate intelligence gathering or pre-operational planning related to 
a cyberattack against digital computer and communication systems and 
networks within the scope of Sec. Sec.  73.54 or 73.110.
    (b) Twenty-four hour recordable events. (1) The licensee shall use 
the site corrective action program to record vulnerabilities, 
weaknesses, failures and deficiencies in their Sec.  73.54 or Sec.  
73.110 cybersecurity program within 24 hours of their discovery.
* * * * *
    (c) * * *
    (6) Declaration of emergencies. Notifications made to the NRC for 
the declaration of an emergency class shall be performed in accordance 
with Sec. Sec.  50.72 or 53.1630 of this chapter, as applicable.
    (7) Elimination of duplication. Separate notifications and reports 
are not required for events that are also reportable under Sec. Sec.  
50.72 and 50.73 or Sec. Sec.  53.1630 and 53.1640 of this chapter. 
However, these notifications should also indicate the applicable Sec.  
73.77 reporting criteria.
* * * * *
0
156. Add Subpart J consisting of Sec. Sec.  73.100 through 73.120 to 
read as follows:
Subpart J--Security Requirements at Commercial Nuclear Plants
Sec.
73.100 Technology-inclusive requirements for physical protection of 
licensed activities at commercial nuclear plants against 
radiological sabotage.
73.110 Technology-inclusive requirements for protection of digital 
computer and communication systems and networks.
73.120 Access authorization program for commercial nuclear plants.

Subpart J--Security Requirements at Commercial Nuclear Plants


Sec.  73.100  Technology-inclusive requirements for physical protection 
of licensed activities at commercial nuclear plants against 
radiological sabotage.

    (a) Introduction. (1) Each licensee that is licensed to operate a 
commercial nuclear plant under 10 CFR part 53 and elects to implement 
the requirements of this section must do so through its physical 
security plan, training and qualification plan, safeguards

[[Page 87120]]

contingency plan, and cybersecurity plan, referred to collectively 
hereafter as ``security plans,'' before initial fuel load into the 
reactor (or, for a fueled manufactured reactor, before initiating the 
physical removal of any one of the independent physical mechanisms to 
prevent criticality required under Sec.  53.620(d)(1) of this chapter).
    (2) The security plans must identify, describe, and account for 
site-specific conditions that affect the licensee's capability to 
satisfy the requirements of this section.
    (b) General performance objective and requirements. (1) The 
licensee must establish, implement, and maintain a physical protection 
program and a security organization, which will have as their objective 
to provide reasonable assurance that activities involving special 
nuclear material are not inimical to the common defense and security 
and do not constitute an unreasonable risk to the public health and 
safety.
    (2) To satisfy the general performance objective of paragraph 
(b)(1) of this section, the physical protection program must protect 
against the design basis threat of radiological sabotage as stated in 
Sec.  73.1. Specifically, the licensee must--
    (i) Ensure that the physical protection program capabilities to 
protect against the design basis threat of radiological sabotage are 
maintained at all times; and
    (ii) Provide defense in depth in achieving performance requirements 
through the integration of engineered systems, administrative controls, 
and management measures.
    (3) The physical protection program must be designed and 
implemented to achieve and maintain the reliability and availability of 
structures, systems, and components (SSCs) required for demonstrating 
compliance with the following performance requirements at all times:
    (i) Intrusion detection. The licensee must be capable of detecting 
attempted and actual unauthorized access to interior and exterior areas 
containing SSCs needed to implement safety and security functions.
    (ii) Intrusion assessment. The licensee must be capable of timely 
assessment for determining the cause of a detected intrusion.
    (iii) Security communication. The licensee must be capable of 
continuous security communications. Communication systems must account 
for design basis threats that can interrupt or interfere with 
continuity or integrity of communications.
    (iv) Security response. The physical protection program must be 
designed to provide timely security response to interdict and 
neutralize adversary attacks up to and including the design basis 
threat of radiological sabotage. The physical protection program must 
be designed to provide layers of security response, with each layer 
assuring that a single failure does not result in the loss of 
capability to neutralize the design basis threat adversary. Structures, 
systems, and components relied on for delay functions must be designed 
to allow for timely security responses to adversary attacks with 
adequate defense in depth.
    (A) The security response may rely on the use of onsite responders, 
law enforcement or other offsite armed responders, or a combination 
thereof, to fulfill the interdiction and neutralization functions 
required by paragraph (b)(3)(iv) of this section. A licensee relying 
entirely or partially on law enforcement or other offsite armed 
responders must--
    (1) Maintain the capability to detect, assess, interdict, and 
neutralize threats as required by paragraphs (b)(3)(i), (b)(3)(ii), and 
(b)(3)(iv) of this section;
    (2) Provide adequate delay to enable law enforcement or other 
offsite armed responders to fulfill the interdiction and neutralization 
functions for threats up to and including the design basis threat of 
radiological sabotage;
    (3) Provide necessary information about the facility and make 
available periodic training to law enforcement or other offsite armed 
responders who will fulfill the interdiction and neutralization 
functions for threats up to and including the design basis threat of 
radiological sabotage;
    (4) Fully describe in the safeguards contingency plan the role that 
law enforcement or other offsite armed responders will play in the 
licensee's protective strategy. The description must provide sufficient 
detail to enable the NRC to determine that the licensee's physical 
protection program provides reasonable assurance of adequate protection 
against threats up to and including the design basis threat of 
radiological sabotage; and
    (5) Identify criteria and measures to compensate for the 
degradation or absence of law enforcement or other offsite armed 
responders and propose suitable compensatory measures that meet the 
requirements of paragraph (h)(3) of this section to address this 
degradation.
    (B) For licensees relying entirely or partially on law enforcement 
responders to fulfill the interdiction and neutralization functions 
required by paragraph (b)(3)(iv) of this section, the training and 
qualification requirements related to armed response personnel in 
paragraphs (c) and (e) of this section do not apply to law enforcement 
responders. The licensee shall continue to satisfy the performance 
evaluation requirements in paragraph (g) of this section for all armed 
response personnel, including law enforcement.
    (v) Protecting against land and waterborne vehicle bomb assaults. 
The licensee must be capable of protecting the plant against the design 
basis threat vehicle bomb assault. The methods that are relied on to 
protect against a design basis threat land vehicle and waterborne 
vehicle bomb assault must be designed to protect the reactor building 
and structures containing safety- or security-related systems, and 
components from explosive effects.
    (vi) Access control portals. The licensee must be capable of 
detecting and denying unauthorized access to persons and pass-through 
of contraband materials (e.g., weapons, incendiaries, explosives) to 
protected areas.
    (4) The licensee must meet the requirements related to target sets 
in Sec.  73.55(f).
    (5) The licensee must identify and analyze site-specific 
conditions, including target sets, that may affect the physical 
protection program needed to implement the requirements of this 
section. The licensee must account for these conditions in 
demonstrating compliance with the requirements of this section.
    (6) The licensee must establish, implement, and maintain a 
performance evaluation program to assess the effectiveness of the 
licensee's implementation of the physical protection program to protect 
against the design basis threat of radiological sabotage.
    (7) The licensee must establish, implement, and maintain an access 
authorization program under Sec.  73.56 and must describe the program 
in the physical security plan.
    (8) The licensee must establish, implement, and maintain a 
cybersecurity program under Sec. Sec.  73.54 or 73.110 and must 
describe the program in the cybersecurity plan.
    (9) The licensee must establish, implement, and maintain an insider 
mitigation program and must describe the program in the physical 
security plan.
    (i) The insider mitigation program must monitor the initial and 
continuing trustworthiness and reliability of individuals granted or 
retaining unescorted access or unescorted access authorization to a 
protected or vital area, and implement defense-in-depth methodologies 
to minimize the potential for an insider (active, passive, or both)

[[Page 87121]]

to adversely affect, either directly or indirectly, the licensee's 
capability to protect against radiological sabotage.
    (ii) The insider mitigation program must integrate elements of--
    (A) The access authorization program under Sec.  73.56;
    (B) The fitness-for-duty program under 10 CFR part 26;
    (C) The cybersecurity program under Sec. Sec.  73.54 or 73.110; and
    (D) The physical protection program under this section.
    (10) The licensee must have the capability to track, trend, 
correct, and prevent recurrence of failures and deficiencies in the 
implementation of the requirements of this section.
    (11) Implementation of security plans and associated procedures 
must be coordinated with other onsite plans and procedures to preclude 
conflict during both normal and emergency conditions and ensure the 
adequate management of the safety and security interface.
    (12)(i) The licensee must ensure that the firearms background check 
requirements of Sec.  73.17 of this part are met for all members of the 
security organization whose official duties require access to covered 
weapons or who inventory enhanced weapons.
    (ii) The provisions of this paragraph are only applicable to 
licensees subject to this section that are also subject to the firearms 
background check provisions of Sec.  73.17 of this part.
    (c) Security organization. The licensee must establish and maintain 
a security organization that is staffed, trained, qualified, and 
equipped to implement the physical protection program under the 
requirements of this section.
    (1) The licensee must establish a management system for maintaining 
and implementing security policies and procedures to implement the 
requirements of this section and the security plans.
    (2) Implementing procedures must document the conduct of security 
operations, security design and configuration controls, maintenance, 
training and qualification, and contingency responses.
    (3) The licensee must--
    (i) Establish a process for the approval of designs, policies, 
processes, and procedures and changes by the individual with overall 
responsibility for the physical protection program; and
    (ii) Ensure that revisions and changes to the physical protection 
program and implementing policies, processes, and procedures satisfy 
the requirements of this section.
    (4) The licensee must retain, in accordance with Sec.  73.70, all 
analyses, assessments, calculations, and descriptions of the technical 
basis for demonstrating compliance with the performance requirements of 
Sec.  73.100(b). The licensee must protect these records in accordance 
with the requirements for protecting safeguards information in 
Sec. Sec.  73.21 and 73.22.
    (5) The licensee may not permit any individual to implement any 
part of the physical protection program unless the individual has been 
trained, equipped, and qualified to perform their assigned duties and 
responsibilities in accordance with the training and qualification 
plan.
    (d) Search requirements. The licensee must establish and implement 
searches of individuals, vehicles, and materials to detect and prevent 
the introduction into the protected area of firearms, explosives, 
incendiary devices, or other items and material which could be used to 
commit radiological sabotage.
    (e) Training and qualification program. The licensee must establish 
and maintain a training and qualification program that ensures 
personnel who are responsible for the physical protection of the 
facility against radiological sabotage are able to effectively perform 
their assigned security-related job duties for implementing the 
requirements of this section and must describe the program in the 
training and qualification plan.
    (f) Security reviews. The licensee must establish and implement 
security reviews to assess the effectiveness of the implementation of 
the physical protection program. Security reviews must be performed by 
individuals independent of those personnel responsible for program 
management and any individual who has direct responsibility for 
implementing the onsite physical protection program.
    (1) The licensee must review each element of the physical 
protection program at a frequency commensurate with the importance or 
significance to safety of plant operations to ensure timely 
identification and documentation of vulnerabilities, improvements, and 
corrective actions. The objective of these reviews must be maintaining 
effective implementation of the engineered and administrative controls 
required to achieve the physical protection program functions and the 
management system required to implement programs and requirements in 
this section.
    (2) The licensee must establish and perform self-assessments to 
ensure the effective implementation of the physical protection program 
functions of detection, assessment, communication, delay, and 
interdiction and neutralization to protect against the design basis 
threat of radiological sabotage. The licensee must perform design 
verification and assessments of the capabilities of active and passive 
engineering systems relied on to protect against the design basis 
threat.
    (3) Reviews of the security program must include, but are not 
limited to, an audit of the effectiveness of the physical protection 
program, security plans, implementing procedures, cybersecurity 
programs, safety/security interface activities, the testing, 
maintenance, and calibration program, and response commitments by 
local, State, and Federal law enforcement authorities.
    (4) The results and recommendations of the onsite physical 
protection program reviews, management's findings regarding program 
effectiveness, and any actions taken as a result of recommendations 
from prior program reviews, must be documented in a report and must be 
maintained in an auditable form and available for inspection.
    (g) Performance evaluation. Licensee performance evaluations must 
include methods appropriate and necessary to assess, test, and 
challenge the integration of the physical protection program's 
functions to protect against the design basis threat, including 
measures to protect against cyberattack and engineered systems designed 
to protect against the design basis threat standalone ground vehicle 
bomb attack.
    (1) The licensee must establish the frequencies for performance 
evaluations commensurate with the security significance of the physical 
protection program.
    (2) The licensee must document processes and procedures for 
implementing the performance evaluations. The licensee must maintain 
records, including results, findings, and corrective actions identified 
during the performance evaluations.
    (h) Maintenance, testing, and calibration and corrective actions. 
(1) The licensee must ensure that security SSCs, including supporting 
systems, are inspected, tested, and calibrated for operability and 
performance at intervals necessary and sufficient to meet the 
requirements of this section.
    (2) The licensee must implement corrective actions to ensure 
resolution of identified vulnerabilities and deficiencies to meet the 
requirements of this section.
    (3) The licensee must establish and implement timely compensatory 
measures for degraded or inoperable security SSCs to meet the 
requirements of this section. Compensatory measures must provide a 
level of protection that is equivalent to the protection that was 
provided prior to the degradation or

[[Page 87122]]

inoperability of the security structures, systems, or components.
    (4) The licensee must document processes and procedures and 
maintain records for implementing the corrective actions, compensatory 
measures, and maintenance, inspection, testing, and calibration of 
security SSCs.
    (i) Suspension of security measures. (1) The licensee may suspend 
implementation of affected requirements of this section in accordance 
with Sec.  53.740(h) of this chapter under the following conditions:
    (i) In an emergency, when action is immediately needed to protect 
the public health and safety; and
    (ii) During severe weather, when the suspension of affected 
security measures is immediately needed to protect the personal health 
and safety of personnel.
    (2) Suspended security measures must be reinstated as soon as 
conditions permit.
    (3) The suspension of security measures must be reported and 
documented in accordance with the provisions of Sec. Sec.  73.1200 and 
73.1205.
    (j) Records. (1) The Commission may inspect, copy, retain, and 
remove all reports, records, and documents required to be kept by 
Commission regulations, orders, or license conditions, whether the 
reports, records, and documents are kept by the licensee or a 
contractor.
    (2) The licensee must maintain all records required to be kept by 
Commission regulations, orders, or license conditions, until the 
Commission terminates the license for which the records were developed 
and must maintain superseded portions of these records for at least 3 
years after the record is superseded, unless otherwise specified by the 
Commission.
    (3) If a contracted security force is used to implement the onsite 
physical protection program, the licensee's written agreement with the 
contractor must be retained by the licensee as a record for the 
duration of the contract.
    (4) Review and audit reports must be available for inspection, for 
a period of 3 years.


Sec.  73.110  Technology-inclusive requirements for protection of 
digital computer and communication systems and networks.

    (a) Each licensee that is licensed to operate a commercial nuclear 
plant under 10 CFR part 53 and elects to implement the requirements of 
this section must establish, implement, and maintain a cybersecurity 
program that is commensurate with the potential consequences resulting 
from cyberattacks, up to and including the design basis threat as 
described in Sec.  73.1. The cybersecurity program must provide 
reasonable assurance that digital computer and communication systems 
and networks are adequately protected against cyberattacks that are 
capable of causing the following consequences:
    (1) Adversely impacting the functions performed by digital assets 
that would prevent a postulated fission product release resulting in 
offsite doses exceeding the values in Sec.  53.210 of this chapter.
    (2) Adversely impacting the functions performed by digital assets 
used by the licensee for implementing the physical security 
requirements in Sec.  53.860(a) of this chapter.
    (b) To protect digital computer and communication systems and 
networks associated with the functions described in paragraphs (a)(1) 
and (2), the licensee must--
    (1) Analyze the potential consequences resulting from cyberattacks 
on digital computer and communication systems and networks and identify 
those assets that must be protected to demonstrate compliance with 
paragraph (a) of this section; and
    (2) Implement the cybersecurity program in accordance with 
paragraph (d) of this section.
    (c) The licensee must comply with the requirements in Sec.  
73.54(a)(2) for the systems and networks identified in paragraph (b)(1) 
of this section in a manner that is commensurate with the potential 
consequences resulting from cyberattacks.
    (d) The cybersecurity program must be designed in a manner that is 
commensurate with the potential consequences resulting from 
cyberattacks through the following steps:
    (1) Implement security controls to protect the assets identified 
under paragraph (b)(1) of this section from cyberattacks, commensurate 
with their safety and security significance;
    (2) Apply and maintain defense-in-depth protective strategies to 
ensure the capability to detect, delay, respond to, and recover from 
cyberattacks capable of causing the consequences identified in 
paragraph (a) of this section;
    (3) Mitigate the adverse effects of cyberattacks capable of causing 
the consequences identified in paragraph (a) of this section; and
    (4) Ensure that the functions of protected assets identified under 
paragraph (b)(1) of this section are not adversely impacted due to 
cyberattacks.
    (e) The licensee must implement the following requirements in a 
manner that is commensurate with the potential consequences resulting 
from cyberattacks:
    (1) As part of the cybersecurity program, the licensee must comply 
with the requirements in Sec.  73.54(d)(1), (2), and (4), and must 
ensure that modifications to assets, identified under paragraph (b)(1) 
of this section are evaluated before implementation to ensure that the 
cybersecurity performance objectives identified in paragraph (a) of 
this section are maintained.
    (2) The licensee must establish, implement, and maintain a 
cybersecurity plan that implements the cybersecurity program 
requirements of this section.
    (i) The cybersecurity plan must describe how the requirements of 
this section will be implemented and must account for the site-specific 
conditions that affect implementation.
    (ii) The cybersecurity plan must include measures for incident 
response and recovery for cyberattacks. The cybersecurity plan must 
include the analysis identified under paragraph (b)(1) of this section 
and describe how the licensee will--
    (A) Apply and maintain defense-in-depth protective strategies as 
required in paragraph (d)(2) of this section;
    (B) Maintain the capability for timely detection and response to 
cyberattacks;
    (C) Mitigate the consequences of cyberattacks;
    (D) Correct exploited vulnerabilities; and
    (E) Restore affected systems, networks, and/or equipment affected 
by cyberattacks.
    (3) The licensee must develop and maintain written policies and 
implementing procedures to implement the cybersecurity plan. Policies, 
implementing procedures, and other supporting technical information 
used by the licensee need not be submitted for Commission review and 
approval as part of the cybersecurity plan but are subject to 
inspection by NRC staff on a periodic basis.
    (4) The licensee must establish and implement cybersecurity reviews 
to assess the effectiveness of the implementation of the cybersecurity 
program.
    (i) The licensee must review each element of the cybersecurity 
program at a frequency commensurate with the importance or significance 
to safety of plant operations to ensure timely identification and 
documentation of vulnerabilities, improvements, and corrective actions.
    (ii) Cybersecurity reviews must be performed by individuals 
independent

[[Page 87123]]

of those personnel responsible for program management and any 
individual who has direct responsibility for implementing the 
cybersecurity program.
    (iii) The licensee must establish and perform self-assessments to 
ensure the effective implementation of the cybersecurity program.
    (iv) The results and recommendations of the cybersecurity program 
reviews, management's findings regarding program effectiveness, and any 
actions taken as a result of recommendations from prior program 
reviews, must be documented in a report and must be maintained in an 
auditable form and available for inspection.
    (5) The licensee must retain all records and supporting technical 
documentation required to demonstrate compliance with the requirements 
of this section as a record until the Commission terminates the license 
for which the records were developed and must maintain superseded 
portions of these records for at least three (3) years after the record 
is superseded, unless otherwise specified by the Commission.


Sec.  73.120  Access authorization program for commercial nuclear 
plants.

    (a) Introduction and scope. Each applicant for an operating license 
or a holder of a combined license under 10 CFR part 53 must establish, 
maintain, and implement an access authorization program before initial 
fuel load into the reactor (or, for a fueled manufactured reactor, 
before initiating the physical removal of any one of the independent 
physical mechanisms to prevent criticality required under Sec.  
53.620(d)(1) of this chapter). The requirements in this section apply 
to licensees satisfying the criterion in Sec.  53.860(a)(2)(i) of this 
chapter.
    (b) Applicability. (1) The following individuals must be subject to 
an access authorization program under this section:
    (i) Any individual to whom a licensee intends to grant unescorted 
access to a commercial nuclear plant protected area, vital area, or 
controlled access area where licensed material is used or stored;
    (ii) Any individual whose duties and responsibilities permit the 
individual to take actions by electronic means, either on site or 
remotely, that could adversely impact the licensee's or applicant's 
operational safety, security, or emergency preparedness;
    (iii) Any individual who has responsibilities for implementing a 
licensee's or applicant's protective strategy, including armed security 
force officers, alarm station operators, and tactical response team 
leaders but not including Federal, State, or local law enforcement 
personnel; and
    (iv) The licensee or applicant access authorization program 
reviewing official or contractor or vendor access authorization program 
reviewers.
    (2) The licensee or applicant may subject other individuals, 
including employees of a contractor or a vendor who are designated in 
access authorization program procedures, to an access authorization 
program that demonstrates compliance with the requirements of this 
section.
    (c) General performance objectives and requirements. Each 
licensee's or applicant's access authorization program under this 
section must demonstrate that the individuals who are specified in 
paragraph (b) of this section are trustworthy and reliable, such that 
they do not constitute an unreasonable risk to public health and safety 
or the common defense and security. The licensee's access authorization 
program must maintain the capabilities for demonstrating compliance 
with the following performance requirements:
    (1) Background investigation. (i)(A) Licensees and applicants must 
ensure that any individual seeking initial unescorted access or to 
maintain unescorted access is subject to a background investigation.
    (B) Background investigations must include the program elements 
contained under Sec.  37.25 of this chapter and must also include a 
credit history evaluation.
    (ii) Background investigations must include fingerprinting and an 
FBI identification and criminal history records check in accordance 
with Sec.  37.27 of this chapter.
    (iii) Licensees must have the informed and signed consent of the 
subject individual to initiate a background investigation. This consent 
must include authorization to share personal information with other 
individuals or organizations as necessary to complete the background 
investigation. A signed consent must be obtained prior to any 
reinvestigation. The subject individual may withdraw his or her consent 
at any time. Licensees must inform the individual that--
    (A) If an individual withdraws his or her consent, the licensee may 
not initiate any elements of the background investigation that were not 
in progress at the time the individual withdrew his or her consent; and
    (B) The withdrawal of consent for the background investigation is 
sufficient cause for denial or termination of unescorted access 
authorization.
    (2) Behavioral observation. Licensees, applicants, contractors, and 
vendors must ensure the access authorization program includes 
provisions that the individuals specified in paragraph (b) of this 
section are subject to behavioral observation.
    (i) Each person subject to behavioral observation must communicate 
to the licensee or applicant observed behaviors or activities of 
individuals that may constitute an unreasonable risk to the health and 
safety of the public and common defense and security.
    (ii) Behavioral observation must include visual observation, in 
person or remotely by video, to detect and promptly report to plant 
supervision any concerns arising from behavioral observation, 
including, but not limited to, concerns related to any questionable 
behavior patterns or activities of others.
    (3) Self-reporting of legal actions. Licensees or applicants must 
inform personnel who are granted and who maintain unescorted access of 
their responsibilities to self-report to plant supervision legal 
actions taken by a law enforcement authority or court of law against 
the individual that could result in incarceration or a court order or 
that requires a court appearance, including but not limited to an 
arrest, an indictment, the filing of charges, or a conviction, but 
excluding minor civil actions or misdemeanors such as parking 
violations or speeding tickets, for any individual who has applied for 
unescorted access or who maintains unescorted access.
    (4) Unescorted access. Licensees or applicants must grant 
unescorted access only after the licensee has verified an individual is 
trustworthy and reliable. A list of persons currently approved for 
unescorted access to a protected area, vital area, or controlled access 
area must be maintained at all times. Unescorted access determinations 
must be reviewed annually by the reviewing official. Licensees and 
applicants must complete an FBI criminal history record check update 
for each individual maintaining unescorted access, within 10 years of 
the last review.
    (5) Termination of unescorted access. Licensees and applicants must 
promptly terminate unescorted access when this access is no longer 
required or a reviewing official determines an individual is no longer 
trustworthy and reliable in accordance with this section.
    (6) Determination basis for access. (i) The licensee's or 
applicant's reviewing official must determine whether to permit, deny, 
unfavorably terminate, maintain, or administratively withdraw an 
individual's unescorted access based on an evaluation of all of the

[[Page 87124]]

information collected to demonstrate compliance with the requirements 
of this section.
    (ii) Licensees and applicants must provide individuals subject to 
this section, prior to any final adverse determination, the right to 
complete, correct, and explain information obtained as a result of the 
licensee's background investigation pursuant to Sec.  37.23(g) of this 
chapter.
    (iii) The licensee's or applicant's reviewing officials are the 
only individuals authorized to make unescorted access determination 
decisions. Each licensee or applicant must name one or more individuals 
to be reviewing officials pursuant to the requirements of Sec.  
37.23(b)(2) of this chapter.
    (7) Review procedures. Review procedures must be established in 
accordance with Sec.  37.23(f) of this chapter, to include provisions 
for the notification in writing of individuals who are denied 
unescorted access or who are unfavorably terminated.
    (8) Protection of information. Licensees, applicants, contractors, 
or vendors must establish and maintain a system of files and procedures 
in accordance with Sec.  37.31 of this chapter, to ensure personal 
information is not disclosed to unauthorized persons.
    (9) Access authorization reviews and corrective action. Licensees 
and applicants must develop, implement, and maintain procedures for 
conduct of access authorization reviews and corrective actions in 
accordance with Sec.  37.33 of this chapter to ensure the continuing 
effectiveness of the access authorization program and to ensure that 
the access authorization program and program elements are in compliance 
with the requirements of this section. Each licensee and applicant must 
be responsible for the continuing effectiveness of the access 
authorization program, including access authorization program elements 
that are provided by the contractors or vendors, and the access 
authorization programs of any of the contractors or vendors that are 
accepted by the licensee or applicant.
    (10) Records. Licensees, applicants, and contractors or vendors 
must document the processes and procedures for maintaining records used 
or created to establish an individual's trustworthiness and reliability 
or to document access determinations. Licensees, applicants, and 
contractor or vendors must--
    (i) Retain documentation regarding the trustworthiness and 
reliability of individual employees for 3 years from the date the 
individual no longer requires unescorted access;
    (ii) Retain a copy of the current access authorization program 
procedures as a record for 3 years after the procedure is no longer 
needed. If any portion of the procedure is superseded, retain the 
superseded material for 3 years after the record is superseded; and
    (iii) Retain the list of persons approved for unescorted access for 
3 years after the list is superseded or replaced. Records maintained in 
any database(s) must be available for NRC review.
0
157. In Sec.  73.1200, revise paragraphs (a) introductory text, (c)(1) 
introductory text, (e)(1) introductory text, (e)(3) and (4), (g)(1) 
introductory text, (o)(5)(i) and (o)(6)(i), (r) and (s) to read as 
follows:


Sec.  73.1200  Notification of physical security events.

    (a) 15-minute notifications--facilities. Each licensee subject to 
the provisions of Sec.  73.20, Sec.  73.45, Sec.  73.46, Sec.  73.51, 
Sec.  73.55, or Sec.  73.100 must notify the NRC Headquarters 
Operations Center, as soon as possible but within 15 minutes after--
* * * * *
    (c) * * *
    (1) Each licensee subject to the provisions of Sec. Sec.  73.20, 
73.45, 73.46, 73.50, 73.51, 73.55, 73.60, 73.67, or 73.100 must notify 
the NRC Headquarters Operations Center as soon as possible but no later 
than 1 hour after the time of discovery of the following significant 
facility security events involving--
* * * * *
    (e) * * *
    (1) Each licensee subject to the provisions of Sec. Sec.  73.20, 
73.45, 73.46, 73.50, 73.51, 73.55, 73.60, 73.67, or 73.100 must notify 
the NRC Headquarters Operations Center within 4 hours after time of 
discovery of the following facility security events involving--
* * * * *
    (3)(i) An event involving a law enforcement response to the 
facility that could reasonably be expected to result in public or media 
inquiries and that does not otherwise require a notification under 
paragraphs (a) through (h) of this section, or in other NRC regulations 
such as Sec.  50.72(b)(2)(xi) or Sec.  53.1630(b)(2)(v) of this 
chapter.
    (ii) As an exemption, licensees need not report law enforcement 
responses to minor incidents, such as traffic accidents.
    (4) For licensees subject to the provisions of Sec.  73.55 or Sec.  
73.100 of this part, an event involving the licensee's suspension of 
security measures.
* * * * *
    (g) * * *
    (1) Each licensee subject to the provisions of Sec.  73.20, Sec.  
73.45, Sec.  73.46, Sec.  73.50, Sec.  73.51, Sec.  73.55, Sec.  73.60, 
Sec.  73.67, or Sec.  73.100 must notify the NRC Headquarters 
Operations Center within 8 hours after time of discovery of the 
following facility security program failures involving--
* * * * *
    (o) * * *
    (5) * * *
    (i) Licensees must establish the requested continuous 
communications channel once the licensee has completed other required 
notifications under this section, Sec.  50.72 of this chapter, appendix 
E to part 50 of this chapter, Sec.  53.1630 of this chapter, Sec.  
70.50 of this chapter; or Sec.  72.75 of this chapter; as appropriate.
* * * * *
    (6) * * *
    (i) Licensees must establish the requested continuous 
communications channel once the licensee or the movement control center 
has completed other required notifications under this section, Sec.  
50.72 of this chapter, appendix E to part 50 of this chapter, Sec.  
53.1630 of this chapter, Sec.  70.50 of this chapter; Sec.  72.75 of 
this chapter; or requested assistance from the LLEA, as appropriate.
* * * * *
    (r) Declaration of emergencies. Licensees notifying the NRC of the 
declaration of an emergency class must do so in accordance with 
Sec. Sec.  50.72, 53.1630, 63.73, 70.50, and 72.75 of this chapter, as 
applicable.
    (s) Elimination of duplication. Licensees with notification 
obligations under paragraphs (a) through (h), (m), and (n) of this 
section and Sec. Sec.  50.72, 53.1630, 63.73, 70.50, and 72.75 of this 
chapter may notify the NRC of events in a single communication. This 
communication must identify each regulation under which the licensee is 
reporting.
* * * * *
0
158. In Sec.  73.1205, revise paragraph (b)(2) to read as follows:


Sec.  73.1205  Written follow-up reports of physical security events.

* * * * *
    (b) * * *
    (2)(i) Licensees subject to Sec.  50.73 or Sec.  53.1640 of this 
chapter must prepare the written follow-up report on NRC Form 366.
    (ii) Licensees not subject to Sec.  50.73 or Sec.  53.1640 of this 
chapter must prepare the written follow-up report in a letter format.
* * * * *

[[Page 87125]]

0
159. In Sec.  73.1210, revise paragraphs (a)(1) and (b)(3)(i) to read 
as follows:


Sec.  73.1210  Recordkeeping of physical security events.

    (a) * * *
    (1) Licensees with facilities or shipment activities subject to the 
provisions of Sec.  73.20, Sec.  73.25, Sec.  73.26, Sec.  73.27, Sec.  
73.37, Sec.  73.45, Sec.  73.46, Sec.  73.50, Sec.  73.51, Sec.  73.55, 
Sec.  73.60, Sec.  73.67, or Sec.  73.100, must record the physical 
security events and conditions adverse to security that are specified 
in paragraphs (c) through (f) of this section.
* * * * *
    (b) * * *
    (3)(i) Licensees must record these physical security events and 
conditions adverse to security in either a stand-alone safeguards event 
log or as part of the licensee's corrective action program, as 
specified under the applicable quality assurance program provisions of 
parts 50, 52, 53, 60, 63, 70, and 72 of this chapter, or both.
* * * * *
0
160. In Sec.  73.1215, revise paragraph (d)(1) introductory text to 
read as follows:


Sec.  73.1215  Suspicious activity reports.

* * * * *
    (d) * * *
    (1) For licensees subject to the provisions of Sec. Sec.  73.20, 
73.45, 73.46, 73.50, 73.51, 73.55, 73.60, 73.67, or 73.100, the 
licensees must report activities they assess are suspicious. Examples 
include, but are not limited to, the following:
* * * * *
0
161. In appendix B to part 73, revise Definitions introductory text to 
read as follows:

Appendix B to Part 73--General Criteria for Security Personnel

* * * * *

Definitions

    Terms defined in parts 50, 53, 70, and 73 of this chapter have 
the same meaning when used in this appendix.
* * * * *

PART 74--MATERIAL CONTROL AND ACCOUNTING OF SPECIAL NUCLEAR 
MATERIAL

0
162. The authority citation for 10 CFR part 74 continues to read as 
follows:

    Authority: Atomic Energy Act of 1954, secs. 53, 57, 161, 182, 
223, 234, 1701 (42 U.S.C. 2073, 2077, 2201, 2232, 2273, 2282, 
2297f); Energy Reorganization Act of 1974, secs. 201, 202 (42 U.S.C. 
5841, 5842); 44 U.S.C. 3504 note.

0
163. In Sec.  74.31, revise paragraph (a) introductory text to read as 
follows:


Sec.  74.31  Nuclear material control and accounting for special 
nuclear material of low strategic significance.

    (a) General performance objectives. Each licensee who is authorized 
to possess and use more than one effective kilogram of special nuclear 
material of low strategic significance, excluding sealed sources, at 
any site or contiguous sites subject to control by the licensee, other 
than a production or utilization facility licensed pursuant to parts 
50, 53, or 70 of this chapter, or operations involved in waste 
disposal, shall implement and maintain a Commission approved material 
control and accounting system that will achieve the following 
objectives:
* * * * *
0
164. In Sec.  74.41, revise paragraph (a) introductory text to read as 
follows:


Sec.  74.41  Nuclear material control and accounting for special 
nuclear material of moderate strategic significance.

    (a) General performance objectives. Each licensee who is authorized 
to possess special nuclear material (SNM) of moderate strategic 
significance or SNM in a quantity exceeding one effective kilogram of 
strategic special nuclear material in irradiated fuel reprocessing 
operations other than as sealed sources and to use this material at any 
site other than a nuclear reactor licensed pursuant to parts 50 or 53 
of this chapter; or as reactor irradiated fuels involved in research, 
development, and evaluation programs in facilities other than 
irradiated fuel reprocessing plants; or an operation involved with 
waste disposal, shall establish, implement, and maintain a Commission-
approved material control and accounting (MC&A) system that will 
achieve the following performance objectives:
* * * * *
0
165. In Sec.  74.51, revise paragraph (a) introductory text to read as 
follows:


Sec.  74.51  Nuclear material control and accounting for strategic 
special nuclear material.

    (a) General performance objectives. Each licensee who is authorized 
to possess five or more formula kilograms of strategic special nuclear 
material (SSNM) and to use such material at any site, other than a 
nuclear reactor licensed pursuant to parts 50 or 53 of this chapter, an 
irradiated fuel reprocessing plant, an operation involved with waste 
disposal, or an independent spent fuel storage facility licensed 
pursuant to part 72 of this chapter shall establish, implement, and 
maintain a Commission-approved material control and accounting (MC&A) 
system that will achieve the following objectives:
* * * * *

PART 75--SAFEGUARDS ON NUCLEAR MATERIAL--IMPLEMENTATION OF 
SAFEGUARDS AGREEMENTS BETWEEN THE UNITED STATES AND THE 
INTERNATIONAL ATOMIC ENERGY AGENCY

0
166. The authority citation for 10 CFR part 75 continues to read as 
follows:

    Authority: Atomic Energy Act of 1954, secs. 53, 63, 103, 104, 
122, 161, 223, 234, 1701 (42 U.S.C. 2073, 2093, 2133, 2134, 2152, 
2201, 2273, 2282, 2297f); Energy Reorganization Act of 1974, sec. 
201 (42 U.S.C. 5841); Nuclear Waste Policy Act of 1982, secs. 135, 
141 (42 U.S.C. 10155, 10161); 44 U.S.C. 3504 note.

0
167. In Sec.  75.4, revise the introductory text and paragraph (6) of 
the definition for ``Facility'' to read as follows:


Sec.  75.4  Definitions.

* * * * *
    Unless otherwise defined in this section, the terms defined in 
Sec. Sec.  40.4, 50.2, 53.020, and 70.4 of this chapter have the same 
meaning when used in this part.
* * * * *
    Facility means:
* * * * *
    (6) Any plant or location where the possession of more than 1 
effective kilogram of nuclear material is licensed pursuant to 10 CFR 
part 40, 50, 53, 60, 61, 63, 70, 72, 76, or 150 of this chapter or an 
Agreement State license.
* * * * *

PART 95--FACILITY SECURITY CLEARANCE AND SAFEGUARDING OF NATIONAL 
SECURITY INFORMATION AND RESTRICTED DATA

0
168. The authority citation for 10 CFR part 95 continues to read as 
follows:

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

0
169. In Sec.  95.5, revise the definition for ``License'' to read as 
follows:


Sec.  95.5  Definitions.

* * * * *
    License means a license issued under 10 CFR part 50, 52, 53, 54, 
60, 63, 70, or 72.
* * * * *

[[Page 87126]]

Sec.  95.39  [Amended]

0
170. In Sec.  95.39(a), remove ``part 52'' and add in its place ``parts 
52 or 53.''

PART 140--FINANCIAL PROTECTION REQUIREMENTS AND INDEMNITY 
AGREEMENTS

0
171. The authority citation for 10 CFR part 140 continues to read as 
follows:

    Authority: Atomic Energy Act of 1954, secs. 161, 170, 223, 234 
(42 U.S.C. 2201, 2210, 2273, 2282); Energy Reorganization Act of 
1974, secs. 201, 202 (42 U.S.C. 5841, 5842); 44 U.S.C. 3504 note.

0
172. In Sec.  140.2, revise paragraphs (a)(1) and (2) to read as 
follows:


Sec.  140.2  Scope.

    (a) * * *
    (1) To each person who is an applicant for or holder of a license 
issued under 10 CFR part 50, 52, 53, or 54 to operate a nuclear 
reactor, and
    (2) With respect to an extraordinary nuclear occurrence, to each 
person who is an applicant for or holder of a license to operate a 
production facility or a utilization facility (including an operating 
license issued under part 50 or part 53 of this chapter and a combined 
license under part 52 or part 53 of this chapter), and to other persons 
indemnified with respect to the involved facilities.
* * * * *
0
173. Revise Sec.  140.10 to read as follows:


Sec.  140.10  Scope.

    This subpart applies to each person who is an applicant for or 
holder of a license issued under 10 CFR parts 50, 53 or 54 to operate a 
nuclear reactor, or is the applicant for or holder of a combined 
license issued under 10 CFR parts 52, 53, or 54, except licenses held 
by persons found by the Commission to be Federal agencies or nonprofit 
educational institutions licensed to conduct educational activities. 
This subpart also applies to persons licensed to possess and use 
plutonium in a plutonium processing and fuel fabrication plant.
0
174. In Sec.  140.11, revise paragraph (b) to read as follows:


Sec.  140.11  Amounts of financial protection for certain reactors.

* * * * *
    (b) In any case where a person is authorized under 10 CFR parts 50, 
52, 53, or 54 to operate two or more nuclear reactors at the same 
location, the total primary financial protection required of the 
licensee for all such reactors is the highest amount which would 
otherwise be required for any one of those reactors; provided, that 
such primary financial protection covers all reactors at the location.
0
175. In Sec.  140.12, revise paragraph (c) to read as follows:


Sec.  140.12  Amount of financial protection required for other 
reactors.

* * * * *
    (c) In any case where a person is authorized under 10 CFR parts 50, 
52, 53, or 54 to operate two or more nuclear reactors at the same 
location, the total financial protection required of the licensee for 
all such reactors is the highest amount which would otherwise be 
required for any one of those reactors; provided, that such financial 
protection covers all reactors at the location.
* * * * *
0
176. Revise Sec.  140.13 to read as follows:


Sec.  140.13  Amount of financial protection required of certain 
holders of construction permits and combined licenses under 10 CFR part 
52.

    Each holder of a 10 CFR part 50 or 10 CFR part 53 construction 
permit, or a holder of a combined license under parts 52 or 53 of this 
chapter before the date that the Commission had made the finding under 
Sec. Sec.  52.103(g) or 53.1452(g) of this chapter, who also holds a 
license under part 70 of this chapter authorizing ownership, possession 
and storage only of special nuclear material at the site of the nuclear 
reactor for use as fuel in operation of the nuclear reactor after 
issuance of either an operating license under 10 CFR part 50 or 53, or 
a combined license under 10 CFR part 52 or 53, shall, during the period 
before issuance of a license authorizing operation under 10 CFR part 50 
or 53, or the period before the Commission makes the finding under 
Sec.  52.103(g) or Sec.  53.1452(g) of this chapter, as applicable, 
have and maintain financial protection in the amount of $1,000,000. 
Proof of financial protection shall be filed with the Commission in the 
manner specified in Sec.  140.15 before issuance of the license under 
part 70 of this chapter.
0
177. In Sec.  140.20, revise paragraphs (a)(1)(i) and (ii) to read as 
follows:


Sec.  140.20  Indemnity agreements and liens.

    (a) * * *
    (1)(i) The effective date of the license (issued under part 50 or 
part 53 of this chapter) authorizing the licensee to operate the 
nuclear reactor involved; or
    (ii) The date that the Commission makes the finding under 
Sec. Sec.  52.103(g) or 53.1452(g) of this chapter; or
* * * * *

PART 150--EXEMPTIONS AND CONTINUED REGULATORY AUTHORITY IN 
AGREEMENT STATES AND IN OFFSHORE WATERS UNDER SECTION 274

0
178. The authority citation for 10 CFR part 150 continues to read as 
follows:

    Authority: Atomic Energy Act of 1954, secs. 11, 53, 81, 83, 84, 
122, 161, 181, 223, 234, 274 (42 U.S.C. 2014, 2201, 2231, 2273, 
2282, 2021); Energy Reorganization Act of 1974, sec. 201 (42 U.S.C. 
5841); Nuclear Waste Policy Act of 1982, secs. 135, 141 (42 U.S.C. 
10155, 10161); 44 U.S.C. 3504 note.

0
179. In Sec.  150.15, revise paragraphs (a)(7)(iii) and (a)(8) to read 
as follows:


Sec.  150.15  Persons not exempt.

    (a) * * *
    (7) * * *
    (iii) Greater than Class C (GTCC) waste, as defined in part 72 of 
this chapter, in an ISFSI or an MRS licensed under part 72 of this 
chapter; the GTCC waste must originate in, or be used by, a facility 
licensed under parts 50, 52, or 53 of this chapter.
    (8) Greater than Class C waste, as defined in part 72 of this 
chapter, that originates in, or is used by, a facility licensed under 
parts 50, 52, or 53 of this chapter and is licensed under part 30 and/
or part 70 of this chapter.
* * * * *

PART 170--FEES FOR FACILITIES, MATERIALS, IMPORT AND EXPORT 
LICENSES, AND OTHER REGULATORY SERVICES UNDER THE ATOMIC ENERGY ACT 
OF 1954, AS AMENDED

0
180. The authority citation for 10 CFR part 170 continues to read as 
follows:

    Authority: Atomic Energy Act of 1954, secs. 11, 161(w) (42 
U.S.C. 2014, 2201(w)); Energy Reorganization Act of 1974, sec. 201 
(42 U.S.C. 5841); 42 U.S.C. 2215; 31 U.S.C. 901, 902, 9701; 44 
U.S.C. 3504 note.

0
181. In Sec.  170.3, revise the definitions for ``Manufacturing 
License,'' ``Part 55 Reviews,'' ``Power reactor,'' and ``Special 
projects'' to read as follows:


Sec.  170.3  Definitions.

* * * * *
    Manufacturing license means a license under subpart F of part 52 of 
this chapter or subpart H of part 53 of this chapter to manufacture a 
nuclear power reactor(s) to be operated at sites not identified in the 
license application.
* * * * *
    Part 55 Reviews as used in this part means those services provided 
by the Commission to administer requalification and replacement 
examinations and tests for reactor

[[Page 87127]]

operators licensed under 10 CFR part 55 or 53 of the Commission's 
regulations and employed by part 50 or 53 licensees. These services 
also include related items such as the preparation, review, and grading 
of the examinations and tests.
* * * * *
    Power reactor means a nuclear reactor designed to produce 
electrical or heat energy licensed by the Commission under the 
authority of section 103 or subsection 104b of the Act, and under the 
provisions of Sec. Sec.  50.21(b), 50.22, or part 53 of this chapter.
* * * * *
    Special projects means specific services provided by the Commission 
for which fees are not otherwise specified in this chapter. This 
includes, but is not limited to, contested hearings on licensing 
actions directly related to U.S. Government national security 
initiatives (as determined by the NRC), topical report reviews, early 
site reviews, waste solidification activities, activities related to 
the tracking and monitoring of shipment of classified matter, services 
provided to certify licensee, vendor, or other private industry 
personnel as instructors for 10 CFR part 55 or 53 reactor operators, 
reviews of financial assurance submittals that do not require a license 
amendment, reviews of responses to Confirmatory Action Letters, reviews 
of uranium recovery licensees' land-use survey reports, and reviews of 
Sec. Sec.  50.71 or 53.1545 of this chapter Final Safety Analysis 
Reports. Special projects does not include activities otherwise exempt 
from fees under this part. It also does not include those contested 
hearings for which a fee exemption is granted in Sec.  170.11(a)(2), 
including those related to individual plant security modifications.
* * * * *
0
182. In Sec.  170.12, revise paragraph (d)(1)(v) to read as follows:


Sec.  170.12  Payment of fees.

* * * * *
    (d) * * *
    (1) * * *
    (v) 10 CFR 50.71 or 53.1545 final safety analysis reports;
* * * * *


Sec.  170.21  [Amended]

0
183. In Sec.  170.21, in footnote 1 remove the phrase ``(e.g., 10 CFR 
50.12, 10 CFR 73.5)'' and add in its place the phrase ``(e.g., 10 CFR 
50.12, 10 CFR 53.080, 10 CFR 73.5)''.
0
184. Revise Sec.  170.41to read as follows:


Sec.  170.41  Failure by an applicant or licensee to pay prescribed 
fees.

    If the Commission determines that an applicant or a licensee has 
failed to pay a prescribed fee required in this part, the Commission 
will not process any application and may suspend or revoke any license 
or approval issued to the applicant or licensee. The Commission may 
issue an order with respect to licensed activities that the Commission 
determines to be appropriate or necessary to carry out the provisions 
of this part, parts 30, 31, 32 through 35, 40, 50, 53, 61, 70, 71, 72, 
73, and 76 of this chapter, and of the Act.

PART 171--ANNUAL FEES FOR REACTOR LICENSES AND FUEL CYCLE LICENSES 
AND MATERIALS LICENSES, INCLUDING HOLDERS OF CERTIFICATES OF 
COMPLIANCE, REGISTRATIONS, AND QUALITY ASSURANCE PROGRAM APPROVALS 
AND GOVERNMENT AGENCIES LICENSED BY THE NRC

0
185. The authority citation for 10 CFR part 171 continues to read as 
follows:

    Authority: Atomic Energy Act of 1954, secs. 11, 161(w), 223, 234 
(42 U.S.C. 2014, 2201(w), 2273, 2282); Energy Reorganization Act of 
1974, sec. 201 (42 U.S.C. 5841); 42 U.S.C. 2215; 44 U.S.C. 3504 
note.

0
186. Revise Sec.  171.3 to read as follows:


Sec.  171.3  Scope.

    The regulations in this part apply to any person holding an 
operating license for a test reactor or research reactor issued under 
part 50 of this chapter, and to any person holding an operating license 
for a power reactor licensed under 10 CFR part 50 or 53, or a combined 
license issued under 10 CFR part 52 or 53, that has provided 
notification to the U.S. Nuclear Regulatory Commission (NRC) that the 
licensee has successfully completed power ascension testing. The 
regulations in this part also apply to any person holding a materials 
license as defined in this part, a Certificate of Compliance, a sealed 
source or device registration, a quality assurance program approval, 
and to a Government agency as defined in this part. Notwithstanding the 
other provisions in this section, the regulations in this part do not 
apply to uranium recovery and fuel facility licensees until after the 
Commission verifies through inspection that the facility has been 
constructed in accordance with the requirements of the license.
0
187. In Sec.  171.5, revise the definitions for ``Operating license,'' 
and ``Power reactor'' to read as follows:


Sec.  171.5  Definitions.

* * * * *
    Operating license means having a license issued under Sec. Sec.  
50.57 or 53.1387 of this chapter. It does not include licenses that 
only authorize possession of special nuclear material after the 
Commission has received a request from the licensee to amend its 
licensee to permanently withdraw its authority to operate or the 
Commission has permanently revoked such authority.
* * * * *
    Power reactor means a nuclear reactor designed to produce 
electrical or heat energy and licensed by the Commission under the 
authority of section 103 or subsection 104b of the Atomic Energy Act of 
1954, as amended, and under the provisions of Sec. Sec.  50.21(b) or 
50.22, or part 53 of this chapter.
* * * * *
0
188. In Sec.  171.15, revise paragraphs (a), (b)(2)(iii), (c)(1), and 
(d)(1) to read as follows:


Sec.  171.15  Annual fees: Non-power production or utilization 
licenses, reactor licenses, and independent spent fuel storage 
licenses.

    (a) Each person holding an operating license for one or more non-
power production or utilization facilities under 10 CFR part 50 that 
has provided notification to the NRC of the successful completion of 
startup testing; each person holding an operating license for a power 
reactor licensed under 10 CFR part 50 or a combined license under 10 
CFR part 52, or an operating license or combined license for a 
commercial nuclear plant under 10 CFR part 53, that has provided 
notification to the NRC of the successful completion of power ascension 
testing; each person holding a 10 CFR part 50 or 52, power reactor 
license, or a 10 CFR part 53 commercial nuclear plant license that is 
in decommissioning or possession only status, except those that have no 
spent fuel onsite; and each person holding a 10 CFR part 72 license who 
does not hold a 10 CFR part 50, 52, or 53 license and provides 
notification under Sec.  72.80(g) of this chapter, shall pay the annual 
fee for each license held during the Federal fiscal year in which the 
fee is due. This paragraph (a) does not apply to test or research 
reactors exempted under Sec.  171.11(b).
    (b) * * *
    (2) * * *
    (iii) Generic activities required largely for NRC to regulate power 
reactors (e.g., updating part 50, part 52, or part 53 of this chapter, 
operating the Incident Response Center, new reactor regulatory 
infrastructure). The base annual fee for operating power reactors does 
not

[[Page 87128]]

include generic activities specifically related to reactor 
decommissioning.
    (c)(1) The FY 2022 annual fee for each power reactor holding a 10 
CFR part 50 operating license or combined license issued under 10 CFR 
part 52 or part 53 that is in a decommissioning or possession-only 
status and has spent fuel onsite, and for each independent spent fuel 
storage 10 CFR part 72 licensee who does not hold a 10 CFR part 50 or 
part 53 operating license, or a 10 CFR part 52 or part 53 combined 
license, is $227,000.
* * * * *
    (d)(1) Each person holding an operating license for an SMR issued 
under 10 CFR part 50 or part 53, or a combined license issued under 10 
CFR part 52 or part 53, that has provided notification to the NRC of 
the successful completion startup testing, shall pay the annual fee for 
all licenses held for an SMR site. The annual fee will be determined 
using the cumulative licensed thermal power rating of all SMR units and 
the bundled unit concept, during the fiscal year in which the fee is 
due. For a given site, the use of the bundled unit concept is 
independent of the number of SMR plants, the number of SMR licenses 
issued, or the sequencing of the SMR licenses that have been issued.
* * * * *
0
189. In Sec.  171.17, revise paragraphs (a) introductory text, 
(a)(1)(ii), and (a)(2) to read as follows:


Sec.  171.17  Proration.

    (a) Reactors, 10 CFR part 72 licensees who do not hold 10 CFR part 
50, 52, or 53 licenses, and materials licenses with annual fees of 
$100,000 or greater for a single fee category. The NRC will base the 
proration of annual fees for terminated and downgraded licenses on the 
fee rule in effect at the time the action is official. The NRC will 
base the determinations on the proration requirements under paragraphs 
(a)(2) and (3) of this section.
    (1) * * *
    (ii) The annual fees for new licenses for non-power production or 
utilization facilities, 10 CFR part 72 licensees who do not hold 10 CFR 
part 50, 52, or 53 licenses, and materials licenses with annual fees of 
$100,000 or greater for a single fee category for the current FY, that 
are subject to fees under this part and are granted a license to 
operate on or after October 1 of a FY, are prorated on the basis of the 
number of days remaining in the FY. Thereafter, the full annual fee is 
due and payable each subsequent FY.
    (2) Terminations. The base operating power reactor annual fee for 
operating reactor licensees or the annual fee for small modular reactor 
licensees, who have requested amendment to withdraw operating authority 
permanently during the FY will be prorated based on the number of days 
during the FY the license was in effect before docketing of the 
certifications for permanent cessation of operations and permanent 
removal of fuel from the reactor vessel or when a final legally 
effective order to permanently cease operations has come into effect. 
The spent fuel storage/reactor decommissioning annual fee for reactor 
licensees who permanently cease operations and have permanently removed 
fuel from the site during the FY will be prorated on the basis of the 
number of days remaining in the FY after docketing of both the 
certifications of permanent cessation of operations and permanent 
removal of fuel from the site. The spent fuel storage/reactor 
decommissioning annual fee will be prorated for those 10 CFR part 72 
licensees who do not hold a 10 CFR part 50, 52, or 53 license who 
request termination of the 10 CFR part 72 license and permanently cease 
activities authorized by the license during the FY based on the number 
of days the license was in effect before receipt of the termination 
request. The annual fee for materials licenses with annual fees of 
$100,000 or greater for a single fee category for the current FY will 
be prorated based on the number of days remaining in the FY when a 
termination request or a request for a possession-only license is 
received by the NRC, provided the licensee permanently ceased licensed 
activities during the specified period. The annual fee for non-power 
production or utilization facilities will be prorated based on the 
number of days remaining in the FY when the authorization to operate 
the facility has been permanently removed from the license during the 
FY.
* * * * *

    Dated: October 7, 2024.

    For the Nuclear Regulatory Commission.
Carrie Safford,
Secretary of the Commission.
[FR Doc. 2024-23434 Filed 10-23-24; 8:45 am]
BILLING CODE 7590-01-P`