86918 Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules 

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. 

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: 

Helen.Chang@nrc.gov. For technical 

questions contact the individuals listed 

in the FOR FURTHER INFORMATION

CONTACT section of this document. 

• Email comments to: 

Rulemaking.Comments@nrc.gov. If you 

do not receive an automatic email reply 

confirming receipt, then contact us at 

301–415–1677. 

• Fax comments to: Secretary, U.S. 

Nuclear Regulatory Commission at 301– 

415–1101. 

• Mail comments to: Secretary, U.S. 

Nuclear Regulatory Commission, 

Washington, DC 20555–0001, ATTN: 

Rulemakings and Adjudications Staff. 

• Hand deliver comments to: 11555 

Rockville Pike, Rockville, Maryland 

20852, between 7:30 a.m. and 4:15 p.m. 

eastern time, Federal workdays; 

telephone: 301–415–1677. 

You can read a plain language 

description of this proposed rule at 

https://www.regulations.gov/docket/ 

NRC-2019-0062. For additional 

direction on obtaining information and 

submitting comments, see ‘‘Obtaining 

Information and Submitting Comments’’ 

in the SUPPLEMENTARY INFORMATION

section of this document. 

FOR FURTHER INFORMATION CONTACT: 

Robert Beall, Office of Nuclear Material 

Safety and Safeguards, telephone: 301– 

415–3874; email: Robert.Beall@nrc.gov; 

or Anders Gilbertson, Office of Nuclear 

Reactor Regulation, telephone: 301– 

415–1541; email: Anders.Gilbertson@

nrc.gov. Both are staff of the U.S. NRC, 

Washington, DC 20555–0001. 

SUPPLEMENTARY INFORMATION: 

Executive Summary 

A. Need for the Regulatory Action 

On January 14, 2019, the President 

signed the Nuclear Energy Innovation 

and Modernization Act (NEIMA) into 

law (Pub. L. 115–439). NEIMA section 

103(a)(4) directs the NRC to ‘‘complete 

a rulemaking to establish a technologyinclusive, regulatory framework for 

optional use by commercial advanced 

nuclear reactor applicants for new 

reactor license applications.’’ NEIMA 

defines a ‘‘technology-inclusive 

regulatory framework’’ as one that is 

‘‘developed using methods of evaluation 

that are flexible and practicable for 

application to a variety of reactor 

technologies, including, where 

appropriate, the use of risk-informed 

and performance-based techniques.’’ 

NEIMA, as further amended by the 

Accelerating Deployment of Versatile, 

Advanced Nuclear for Clean Energy Act 

of 2024 (ADVANCE Act), defines the 

term ‘‘advanced nuclear reactor’’ as ‘‘a 

nuclear fission reactor or fusion 

machine, including a prototype plant (as 

defined in sections 50.2 and 52.1 of title 

10, Code of Federal Regulations (as in 

effect on the date of enactment of 

[NEIMA])), with significant 

improvements compared to commercial 

nuclear reactors under construction as 

of the date of enactment of [NEIMA].’’ 

The NRC initially considered 

establishing the scope of proposed part 

53, ‘‘Risk-Informed, TechnologyInclusive Regulatory Framework for 

Commercial Nuclear Plants,’’ of title 10 

of the Code of Federal Regulations (10 

CFR) as being for ‘‘advanced nuclear 

plants’’ consisting of one or more 

‘‘advanced nuclear reactors’’ as defined 

in NEIMA. Based on public discussions 

on the use of the term, the NRC 

determined that the NEIMA definition, 

although broad, did not define 

‘‘significant improvements’’ with 

enough specificity to implement in NRC 

regulations. Additionally, a number of 

stakeholders suggested that the 

descriptor, ‘‘advanced,’’ implied 

enhanced safety, while the NEIMA 

definition includes ‘‘significant 

improvements’’ in areas other than 

safety enhancements. In response to this 

feedback, and to be technology 

inclusive, the NRC determined that the 

broader term ‘‘commercial nuclear 

plant’’ would be preferable. 

The current application and licensing 

requirements in 10 CFR part 50, 

‘‘Domestic Licensing of Production and 

Utilization Facilities,’’ and 10 CFR part 

52, ‘‘Licenses, Certifications, and 

Approvals for Nuclear Power Plants,’’ 

were primarily developed to address 

license requests concerning watercooled reactors, and to address 

operational requirements for those types 

of reactors. This proposed rule responds 

to NEIMA by creating an alternative 

regulatory framework for licensing 

future commercial nuclear plants. The 

new alternative requirements and 

implementing guidance would adopt 

technology-inclusive approaches and 

use risk-informed and performancebased techniques to ensure an 

equivalent level of safety to that of 

operating commercial nuclear plants 

while providing flexibility for licensing 

and regulating a variety of technologies 

and designs for commercial nuclear 

reactors. 

B. Major Provisions 

Major provisions of this proposed 

rule, supported by accompanying 

guidance, include the following: 

• A new alternative technologyinclusive, risk-informed, performancebased framework that includes 

requirements for licensing and 

regulating nuclear plants during the 

various stages of their life cycles. 

• A new alternative technologyinclusive, risk-informed, and 

performance-based framework in 10 

CFR part 26, ‘‘Fitness for Duty 

Programs,’’ developed from existing 

requirements in subpart K, ‘‘FFD 

Programs for Construction,’’ of part 26. 

• A new alternative technologyinclusive and performance-based 

security framework in 10 CFR part 73, 

‘‘Physical Protection of Plants and 

Materials,’’ that includes requirements 

for protection of licensed activities at 

commercial nuclear plants. 

C. Costs and Benefits 

The NRC prepared a draft regulatory 

analysis to determine the expected 

quantitative costs and benefits of this 

proposed rule and associated guidance 

as well as qualitative factors to be 

considered in the NRC’s rulemaking 

decision. The conclusion from the 


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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 

PDR.Resource@nrc.gov. 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 PDR.Resource@nrc.gov 

or call 1–800–397–4209 or 301–415– 

4737, between 8 a.m. and 4 p.m. eastern 

time, Monday through Friday, except 

Federal holidays. 

B. Submitting Comments 

The NRC encourages electronic 

comment submission through the 

Federal rulemaking website (https://

www.regulations.gov). Please include 

Docket ID NRC–2019–0062 in your 

comment submission. To facilitate NRC 

review, please distinguish between 

comments on the proposed rule and 

comments on the proposed guidance. 

The NRC cautions you not to include 

identifying or contact information that 

you do not want to be publicly 

disclosed in your comment submission. 

The NRC will post all comment 

submissions at https://

www.regulations.gov as well as enter the 

comment submissions into ADAMS. 

The NRC does not routinely edit 

comment submissions to remove 

identifying or contact information. 

If you are requesting or aggregating 

comments from other persons for 

submission to the NRC, then you should 

inform those persons not to include 

identifying or contact information that 

they do not want to be publicly 

disclosed in their comment submission. 

Your request should state that the NRC 

does not routinely edit comment 

submissions to remove such information 

before making the comment 

submissions available to the public or 

entering the comment into ADAMS. 

II. Background 

A. NRC Advanced Reactor Readiness 

In its ‘‘Policy Statement on the 

Regulation of Advanced Nuclear Power 

Plants,’’ dated July 8, 1986, the 

Commission stated that it considered 

the term ‘‘advanced’’ to apply to 

reactors that are significantly different 

from current (i.e., current in 1986) 

generation light-water reactors (LWRs) 

then under construction or in operation, 

and that ‘‘advanced’’ includes reactors 

that provide enhanced margins of safety 

or utilize simplified inherent or other 

innovative means to accomplish their 

safety functions. At the time, certain 

high temperature gas-cooled reactors, 

liquid metal reactors, and LWRs of 

innovative design were considered to be 

‘‘advanced.’’ The 1986 policy statement 


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provided the Commission’s policy 

regarding the review of, and desired 

characteristics associated with, 

advanced reactors. The NRC updated 

this statement in the ‘‘Policy Statement 

on the Regulation of Advanced 

Reactors,’’ dated October 14, 2008 

(Advanced Reactor Policy Statement). 

The agency has undertaken many 

activities related to advanced reactors, 

including issuing an advance notice of 

proposed rulemaking titled, 

‘‘Approaches to Risk-Informed and 

Performance-Based Requirements for 

Nuclear Power Reactors,’’ dated May 4, 

2006 (71 FR 26267). These efforts were 

often done in parallel, and sometimes 

interwoven, with the NRC’s efforts to 

improve risk-informed and 

performance-based approaches within 

the agency (e.g., the Commission’s 

policy statement, ‘‘Use of Probabilistic 

Risk Assessment Methods in Nuclear 

Regulatory Activities,’’ dated August 16, 

1995 (PRA Policy Statement)). 

In 2016, the NRC issued ‘‘NRC Vision 

and Strategy: Safely Achieving Effective 

and Efficient Non-Light-Water Mission 

Readiness’’ (Advanced Reactor Vision 

and Strategy Document), in response to 

increasing interest in advanced reactor 

designs. The NRC considered the 

Department of Energy’s (DOE’s) 

advanced reactor deployment goals in 

developing the Advanced Reactor 

Vision and Strategy Document. Since 

publication of the document, the NRC 

continues to manage its activities to 

support the DOE’s deployment goals. 

The Advanced Reactor Vision and 

Strategy Document identified initiating 

and developing a new risk-informed and 

performance-based regulatory 

framework as a possible long-term goal. 

However, the NRC staff’s initial efforts 

were focused on resolving policy issues 

and developing guidance for licensing 

non-LWR technologies under the 

existing regulatory frameworks (parts 50 

and 52). The NRC staff issues annual 

Commission papers on the status and 

progress of the NRC staff’s activities 

related to advanced reactors (e.g., 

SECY–24–0020, ‘‘Advanced Reactor 

Program Status,’’ dated February 27, 

2024). These Commission papers 

provide status updates for advanced 

reactor activities undertaken both prior 

to and after initiation of this 

rulemaking. 

In 2017, the NRC staff prioritized 

activities to support the development of 

technology-inclusive, risk-informed, 

and performance-based licensing 

approaches that could be implemented 

under the existing regulatory framework 

in parts 50 and 52. One key element of 

these efforts was the Licensing 

Modernization Project (LMP), a costshared initiative led by nuclear utilities 

and supported by DOE. The LMP is a 

technology-inclusive, risk-informed, 

and performance-based methodology 

developed for non-LWR designs. The 

LMP provides a systematic and 

reproducible process for licensing-basis 

event (LBE) selection and evaluation; 

classification of structures, systems, and 

components (SSCs); and assessment of 

defense in depth. The LMP refined the 

DOE’s Next Generation Nuclear Plant 

Program methodologies to reflect 

interactions with the NRC, to address 

feedback from industry, and to broaden 

the scope of the approach to ensure 

applicability to various non-LWR 

technologies. The LMP activities led to 

the publication and submittal of Nuclear 

Energy Institute (NEI) 18–04, Revision 1, 

‘‘Risk-Informed Performance-Based 

Technology Inclusive Guidance for NonLight Water Reactor Licensing Basis 

Development,’’ issued August 2019. The 

document indicates that controlling the 

frequencies and potential consequences 

of a wide spectrum of events is the 

primary focus of the LMP approach. 

The NRC endorsed the principles and 

methodology in NEI 18–04, with 

clarifications, in RG 1.233, ‘‘Guidance 

for a Technology-Inclusive, RiskInformed, and Performance-Based 

Methodology to Inform the Licensing 

Basis and Content of Applications for 

Licenses, Certifications, and Approvals 

for Non-Light-Water Reactors.’’ The 

NRC staff sought Commission approval 

of the use of LMP and NEI–18–04 in 

SECY–19–0117, ‘‘Technology-Inclusive, 

Risk-Informed, and Performance-Based 

Methodology to Inform the Licensing 

Basis and Content of Applications for 

Licenses, Certifications, and Approvals 

for Non-Light-Water Reactors,’’ dated 

December 2, 2019. In that paper, the 

staff described the relationship between 

the LMP and NEI–18–04 and previous 

relevant Commission decisions, 

including those described in SECY–93– 

092, ‘‘Issues Pertaining to the Advanced 

Reactor (PRISM, MHTGR, and PIUS) 

and CANDU 3 Designs and their 

Relationship to Current Regulatory 

Requirements,’’ dated April 8, 1993. The 

Commission approved the use of the 

LMP methodology and NEI–18–04 as a 

reasonable approach for establishing key 

parts of the licensing basis and content 

of applications for licenses, 

certifications, and approvals for nonLWRs in Staff Requirements 

Memorandum (SRM) SRM–SECY–19– 

0117, dated May 26, 2020. Although the 

LMP approach is technology- inclusive, 

the industry and NRC staff initially 

focused the LMP’s applicability on nonLWRs, both for efficiency and to support 

near-term non-LWR applications under 

the existing regulatory framework, such 

as the Advanced Reactor Demonstration 

Projects supported by DOE. 

As stated in the part 53 rulemaking 

plan, SECY–20–0032, the NRC staff 

developed part 53 by building upon 

recent and ongoing activities such as the 

LMP approach described in SECY–19– 

0117. Such an approach supports 

implementing the NEIMA requirement 

to use, where appropriate, risk-informed 

and performance-based techniques, and 

it also capitalizes on previous initiatives 

by the industry, DOE, and the NRC, 

including the LMP. This approach 

highlights the role of PRA in riskinformed and performance-based 

approaches to identifying enhanced 

safety margins that can be used to justify 

operational flexibilities. The proposed 

framework is largely based on the 

methodology described in SECY–19– 

0117 and includes a prominent role for 

PRA. 

As discussed in section II.B, 

‘‘Stakeholder Views on Part 53 

Preliminary Proposed Rule Language,’’ 

of this document, the NRC conducted 

extensive public outreach on early 

versions of the proposed rule text. Early 

versions of the draft proposed rule 

included two alternative regulatory 

frameworks. One framework (called 

‘‘Framework A’’) offered a licensing 

approach centered largely on risk 

analysis and the other framework 

(called ‘‘Framework B’’) largely 

replicated the existing licensing 

approach in parts 50 and 52 but 

modified it to be technology neutral. In 

its SRM to SECY–23–0021, ‘‘Proposed 

Rule: Risk-Informed, TechnologyInclusive Regulatory Framework for 

Advanced Reactors (RIN 3150–AK31),’’ 

the Commission disapproved the 

inclusion of Framework B in this 

proposed rule and directed the staff to 

provide them within one year an 

options paper for possible future use of 

the Framework B methodology. 

B. Stakeholder Views on Part 53 

Preliminary Proposed Rule Language 

In SRM–SECY–20–0032, the 

Commission directed the NRC staff to 

prepare and release preliminary 

proposed rule language, followed by 

public outreach and dialogue, and then 

further revise the language until the 

NRC staff had established the rudiments 

of its proposed rule for Commission 

consideration. To implement the 

Commission’s direction, the NRC staff 

undertook an unprecedented program of 

stakeholder engagement, recognizing the 

importance of this rulemaking to the 

advanced reactor community and 


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interested stakeholders from a broad 

range of backgrounds and organizations. 

On November 6, 2020, the NRC 

published a notification in the Federal 

Register (85 FR 71002) describing plans 

for the periodic release of preliminary 

proposed rule language, meetings with 

stakeholders, and the ability of 

stakeholders to provide input during the 

development of this proposed rule. 

Sections of the preliminary proposed 

rule language were subsequently 

released, and the NRC held numerous 

public meetings to discuss the 

preliminary proposed rule language and 

obtain input from stakeholders. On 

December 10, 2021, the NRC published 

a second notification in the Federal 

Register (86 FR 70423) announcing that 

the development of the proposed rule 

and related interactions with 

stakeholders were being extended until 

August 31, 2022. 

By the close of the public stakeholder 

interactions on August 31, 2022, the 

NRC staff had held 24 public meetings 

since September 2020. The NRC staff 

also met with the Advisory Committee 

on Reactor Safeguards (ACRS) in 16 

public meetings during this period. By 

the close of the public engagement 

period on the preliminary proposed rule 

language, 126 letters were received on 

the preliminary proposed rule language. 

Of these 126 letters, 21 were from nongovernmental organizations, 31 were 

from the public, one was from Congress, 

and the remaining 73 letters were from 

NRC licensees, the NEI, and other 

industry groups. In addition, the ACRS 

wrote four interim letter reports to the 

Chair on this rulemaking and issued its 

final letter report on November 22, 

2022. The letters from stakeholders 

provided various points of view and 

suggestions for clarifications, additions, 

and deletions to the preliminary 

proposed rule language. Copies of these 

letters may be viewed and downloaded 

from the Federal rulemaking website 

https://www.regulations.gov, under 

docket number NRC–2019–0062. The 

inputs received were considered in the 

development of this proposed rule. 

However, as described during the 

various public interactions related to 

this rulemaking and in supporting 

documents, the NRC will not formally 

disposition the questions and 

suggestions related to the preliminary 

proposed rule language as it will for 

public comments received following the 

publication of this proposed rule. 

III. Discussion 

A. Objective and Applicability 

The NRC is proposing to add a new, 

alternative part to its regulations that 

would set out a risk-informed, 

technology-inclusive framework for the 

licensing and regulation of commercial 

nuclear plants. This new approach 

would achieve the following: (1) 

continue to provide reasonable 

assurance of adequate protection of 

public health and safety and the 

common defense and security; (2) 

promote regulatory stability, 

predictability, and clarity; (3) reduce 

requests for exemptions from the 

current requirements in parts 50 and 52; 

(4) establish new requirements to 

address non-LWR technologies; (5) 

recognize technological advancements 

in reactor design; and (6) credit the 

possible response of some designs of 

commercial nuclear plants to postulated 

accidents, including slower transient 

response times and relatively small and 

slow release of fission products. This 

proposed rule would add 10 CFR part 

53; subpart M, ‘‘Fitness for Duty 

Programs for Facilities Licensed Under 

10 CFR Part 53,’’ to Part 26; § 73.100, 

‘‘Technology-inclusive requirements for 

physical protection of licensed activities 

at commercial nuclear plants against 

radiological sabotage,’’ § 73.110, 

‘‘Technology-inclusive requirements for 

protection of digital computer and 

communication systems and networks,’’ 

and § 73.120, ‘‘Access authorization 

program for commercial nuclear 

plants,’’ as well as make conforming 

changes throughout 10 CFR chapter I, 

‘‘Nuclear Regulatory Commission.’’ 

B. Need for Changes to the Existing 

Regulatory Framework 

The NRC has long recognized that the 

licensing and regulation of a variety of 

nuclear reactor technologies would 

present challenges because the existing 

regulatory framework has evolved 

primarily to address the LWR designs 

that compose the current operating fleet 

(widely referred to as Generation II 

reactors). The NRC has had many 

interactions with designers of various 

reactor technologies under 

development, sometimes collectively 

referred to as advanced reactors (widely 

referred to as Generation III/III+ (i.e., 

evolutionary light-water) and 

Generation IV (i.e., non-light-water) 

reactors). The interactions have 

informed the development of policies 

and guidance to support the potential 

licensing of new and different types of 

reactor facilities, some of which may not 

utilize LWR designs. The NRC issued its 

Advanced Reactor Policy Statement to 

provide all interested parties, including 

the public, with the Commission’s 

views concerning the desired 

characteristics of advanced reactor 

designs. The NRC further described its 

early efforts to establish a technologyinclusive approach to the regulation of 

nuclear reactors in the advance notice of 

proposed rulemaking published in 2006. 

The NRC acknowledged in its ‘‘Report 

to Congress: Advanced Reactor 

Licensing,’’ issued August 2012, that 

while the safety philosophy inherent in 

the current regulations applies to all 

reactor technologies, the specific and 

prescriptive aspects of those regulations 

clearly focus on the current fleet of LWR 

facilities. 

Congress similarly recognized the 

potential benefits of developing a 

regulatory infrastructure to support the 

development and commercialization of 

advanced nuclear reactors. 

Consequently, Congress passed NEIMA 

in late 2018, and the President signed it 

into law in January 2019. NEIMA 

directed the NRC to undertake a 

rulemaking to establish a technologyinclusive regulatory framework for 

optional use by applicants for new 

commercial advanced nuclear reactor 

licenses. In addition, on July 9, 2024, 

the President signed into law the 

Accelerating Deployment of Versatile, 

Advanced Nuclear for Clean Energy Act 

of 2024, also referred to as the 

ADVANCE Act. The NRC is evaluating 

its plans for implementing the 

ADVANCE Act, including how its 

regulations, as well as the proposed part 

53 or future revisions to it, could be 

used to address provisions in the 

ADVANCE Act. The ADVANCE Act 

contains provisions on a variety of 

nuclear-related topics, such as micro 

reactors, nuclear reactor license 

application reviews, and nuclear fuel. In 

Section VI, ‘‘Specific Requests for 

Comments,’’ the NRC is requesting 

public input on how part 53 could be 

revised to better enable its potential use 

to implement the ADVANCE Act. 

The requirements in part 53 would 

support a wide variety of potential 

commercial nuclear reactor 

technologies. As noted in this 

discussion, the current regulatory 

framework in parts 50 and 52 evolved 

in the context of the current operating 

reactor fleet dominated by LWRs and as 

a result includes provisions specific to 

LWR technologies. While the NRC can 

license other reactor technologies under 

the current framework by using existing 

regulatory flexibilities and the 

exemption process, there is significant 

interest in developing a regulatory 

framework that is flexible enough to 

accommodate multiple technologies and 

robust enough to ensure a level of safety 

equivalent to parts 50 and 52, consistent 

with the Commission’s Advanced 

Reactor Policy Statement. The 

Commission reiterated its safety 


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expectations for new reactors in the 

SRM for SECY–10–0121, ‘‘Modifying 

the Risk-Informed Regulatory Guidance 

for New Reactors,’’ dated March 2, 2011: 

Because new plant designs incorporate 

operating experience from current generation 

reactors, severe accident research, and risk 

insights from design probabilistic risk 

assessments, the Commission expects that 

the advanced technologies incorporated in 

new reactors will result in enhanced margins 

of safety. However, the Commission 

continues to expect (consistent with the 2008 

Advanced Reactor Policy Statement), as a 

minimum, at least the same degree of 

protection of the public and the environment 

that is required for current-generation lightwater reactors. New reactors with these 

enhanced margins and safety features should 

have greater operational flexibility than 

current reactors. 

However, developing a regulatory 

framework that can accommodate a 

wide range of technologies while 

maintaining an acceptable level of safety 

presents significant regulatory 

challenges. The existing regulations 

have been developed over the course of 

decades and reflect changes to address 

events discovered through operating 

experience. In contrast, part 53 is being 

developed to accommodate technologies 

that, in some cases, lack significant 

operating experience. To address these 

challenges, the NRC drew on welldeveloped approaches to licensing to 

produce a technology-neutral and robust 

regulatory framework. The proposed 

regulatory framework would use PRAs 

to assess risks, help establish technical 

requirements, and manage operations. 

The framework builds on the LMP, 

which is a technology-inclusive 

approach to licensing that leverages 

insights from a detailed PRA to provide 

applicants with significant design and 

operation flexibilities. 

C. 10 CFR Part 53: Framework 

This proposed rule consists of several 

major components, including a new part 

53, to be added to 10 CFR chapter I, 

revisions for part 26, part 50, and part 

73, and conforming changes throughout 

10 CFR chapter I. 

Part 53 is comprised of subparts A 

through M. These provisions are 

organized to provide high-level 

performance criteria and to specify 

requirements to demonstrate 

compliance with those performance 

criteria throughout major stages of the 

life cycle of commercial nuclear plants. 

This organization reflects a systemsengineering style approach to the 

design, licensing, operation, and 

ultimately decommissioning of future 

commercial nuclear plants. Organizing 

requirements in this manner also 

supports performance-based 

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, §§ 50.1 through 

50.13) but would reference the 

corresponding regulations in part 53 in 

place of references to part 50. 

A. Discussion of Definitions in Proposed 

Part 53 

This proposed rule would include a 

definition section in § 53.020. The 

definitions of most terms in § 53.020 

would be equivalent to the 

corresponding terms defined in: (1) 

§§ 50.2, 52.1, and other NRC 

regulations; (2) NEI 18–04, as endorsed 

by RG 1.233; or (3) American Society of 

Mechanical Engineers (ASME)/ 

American Nuclear Society Risk 

Assessment Standard (RA–S)-1.4–2021, 

as endorsed for trial use by RG 1.247, 

‘‘Acceptability of Probabilistic Risk 

Assessment Results for Non-Light-Water 

Reactor Risk-Informed Activities.’’ This 

is intended to provide clarity and 

consistency in terminology where 

possible and to utilize past and ongoing 

NRC initiatives to support the licensing 

of new reactors. Specific deviations 

from existing definitions are further 

explained in the following paragraphs. 

Regarding the definition of 

‘‘Commercial nuclear plant’’ and 

‘‘Commercial nuclear reactor’’ in 

proposed § 53.020, as noted previously, 

the NRC initially considered 

establishing the scope of part 53 as 

being for ‘‘advanced nuclear plants.’’ 

The preliminary proposed rule language 

defined ‘‘advanced nuclear plant’’ as ‘‘a 

utilization facility consisting of one or 

more advanced nuclear reactors’’ as 

defined in NEIMA. NEIMA defines the 

term ‘‘advanced nuclear reactor’’ as ‘‘a 


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nuclear fission reactor or fusion 

machine, including a prototype plant (as 

defined in sections 50.2 and 52.1 of title 

10, Code of Federal Regulations (as in 

effect on the date of enactment of this 

Act)), with significant improvements 

compared to commercial nuclear 

reactors under construction as of the 

date of enactment of this Act, including 

improvements such as—(A) additional 

inherent safety features; (B) significantly 

lower levelized cost of electricity; (C) 

lower waste yields; (D) greater fuel 

utilization; (E) enhanced reliability; (F) 

increased proliferation resistance; (G) 

increased thermal efficiency; or (H) 

ability to integrate into electric and 

nonelectric applications.’’ 

Based on public discussions on the 

use of the term, the NRC determined 

that the NEIMA definition, although 

broad, did not define ‘‘significant 

improvements’’ with enough specificity 

to implement in NRC regulations. 

Additionally, a number of stakeholders 

suggested that the descriptor, 

‘‘advanced,’’ implied enhanced safety, 

while the NEIMA definition includes 

‘‘significant improvements’’ in areas 

other than safety enhancements. In 

response to this feedback, and to be 

technology inclusive, the NRC 

determined that the broader term 

‘‘commercial nuclear plant’’ would be 

preferable. The NEIMA definition of 

advanced nuclear reactor also includes 

fusion technologies. Fusion energy 

systems have not been included in the 

scope of part 53 but are the subject of 

a separate rulemaking activity, 

‘‘Regulatory Framework for Fusion 

Systems.’’ See NRC docket ID NRC– 

2023–0017 on the Federal rulemaking 

website http://www.regulations.gov. 

The NRC proposes to allow use of part 

53 by any ‘‘commercial nuclear plant.’’ 

The use of the term ‘‘plant’’ versus 

‘‘reactor,’’ as used in existing 

regulations (i.e., § 50.2), recognizes that 

co-located support facilities and 

radionuclide sources need to be 

considered in the licensing of a facility. 

The phrase ‘‘commercial purposes,’’ as 

used in the definition of ‘‘commercial 

nuclear plant,’’ includes purposes such 

as providing process heat for a variety 

of industrial applications (e.g., 

desalination, oil refining, hydrogen 

production). The NRC has not compiled 

a complete list of such commercial 

purposes. The definition of 

‘‘Commercial nuclear plant’’ refers to a 

‘‘Commercial nuclear reactor,’’ which is 

defined based on the definition of 

‘‘Nuclear reactor’’ in § 50.2. However, 

the phrase ‘‘in a self-supporting chain 

reaction’’ was removed from the 

definition to enable applying part 53 to 

accelerator driven systems that use 

special nuclear material (SNM) but that 

do not involve self-sustaining chain 

reactions. Relatedly, ‘‘Utilization 

facility’’ is also defined in § 53.020 

based on the definition of that term in 

§ 50.2 but is also revised to refer to a 

‘‘Commercial nuclear plant’’ as defined 

in § 53.020. 

The NRC proposes to include a 

definition of ‘‘Consensus code or 

standard’’ in part 53 that is based on the 

use of these terms in the National 

Technology Transfer and Advancement 

Act of 1995 (NTTAA) (Pub. L. 104–113) 

and the Office of Management and 

Budget (OMB) Circular No. A–119, 

‘‘Federal Participation in the 

Development and Use of Voluntary 

Consensus Standards and in Conformity 

Assessment Activities.’’ As required by 

NTTAA, the NRC undertakes the 

following activities: (i) consults with 

voluntary consensus standards bodies; 

(ii) participates with voluntary 

consensus bodies in the development of 

consensus standards; and (iii) uses 

consensus standards as a means to carry 

out the NRC’s policy objectives. In part 

53, the NRC is not proposing to 

incorporate by reference specific codes 

and standards as is done under the 

existing regulations in § 50.55a, ‘‘Codes 

and standards,’’ because some codes 

and standards are LWR-specific. Part 53 

would require that design features must 

be designed using generally accepted 

consensus codes and standards but 

would not incorporate the specific code 

or standard into the NRC’s regulations. 

During public meetings, significant 

discussions with stakeholders indicated 

that future reactor designers were 

interested in the use of international 

consensus standards that have not yet 

been endorsed by the NRC. The 

definition proposed in part 53 would 

allow for the use of international codes 

and standards not previously used in 

NRC licensing but recognizes that the 

use of any consensus code or standard 

would ultimately need to be found 

acceptable by the NRC, either through 

generic efforts to endorse a code or 

standard or on an application-specific 

basis during an individual licensing 

review. 

The proposed definition of 

‘‘Construction’’ is slightly different than 

the definition in § 50.10—it would cover 

the same concept but be applied to a 

slightly different scope of activities 

based on how SSCs are classified under 

part 53. In part 53, the definition of 

‘‘Construction’’ is based on the 

definition in § 50.10 but modified to 

apply to safety-related (SR) and nonsafety-related but safety-significant 

(NSRSS) SSCs identified by the design 

and analysis requirements in subparts B 

and C to ensure the safety criteria are 

met. 

Section 53.020 would also add 

definitions for terms related to event 

selection (LBEs, design-basis accidents 

(DBAs), anticipated event sequences, 

unlikely event sequences, and very 

unlikely event sequences); equipment 

classifications (SR, NSRSS, and nonsafety-significant SSCs); performance 

metrics (e.g., safety criteria and 

functional design criteria); and special 

treatment. 

The regulation would define ‘‘Safety 

criteria’’ in terms of the plant-level 

performance-based metrics that would 

be provided in §§ 53.210 and 53.220. 

The term ‘‘Functional design criteria’’ 

would be defined as metrics for the 

performance of specific SSCs that are 

determined from the role of the SSC in 

meeting the safety criteria. These are 

new terms that have not previously been 

defined or used in NRC regulation. 

The term ‘‘Safety-related SSCs’’ 

would refer to those SSCs needed to 

meet the safety criteria in § 53.210. The 

term ‘‘Non-safety-related but safetysignificant SSCs’’ would mean those 

SSCs that are not SR because they are 

not relied upon to perform any function 

necessary to demonstrate compliance 

with § 53.210 but warrant special 

treatment because they are relied on to 

achieve adequate defense in depth or 

perform risk-significant functions. The 

term ‘‘Special treatment’’ would be 

defined as requirements, such as quality 

assurance and programmatic controls, 

identified for each design feature to 

ensure that the safety criteria are 

satisfied and the safety functions are 

fulfilled. These requirements would also 

ensure that SR and NSRSS SSCs will 

provide defense in depth, or perform 

risk-significant functions, under service 

conditions and with SSC reliabilities 

that are consistent with the analysis 

required in proposed subpart C. 

Structures, systems, and components 

designated as SR would also contribute 

to defense in depth and risk-significant 

functions and may warrant special 

treatments beyond those defined for the 

SR functions needed for compliance 

with § 53.210. The term ‘‘Non-safetysignificant SSCs’’ would mean those 

SSCs that are not SR or NSRSS. 

The terms ‘‘Design-basis accidents,’’ 

‘‘Anticipated event sequences,’’ 

‘‘Unlikely event sequences,’’ and ‘‘Very 

unlikely event sequences’’ would be 

defined to be different types of 

‘‘Licensing-basis events’’ and would also 

be largely equivalent to the LMP’s 

definitions of DBAs, anticipated 

operational occurrences (AOOs), designbasis events (DBEs), and beyond-designbasis events, respectively. The term 


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‘‘Design-basis accidents’’ would be 

defined as postulated event sequences 

that are used to set functional design 

criteria and performance objectives for 

the design of SR SSCs through 

deterministic analyses. Design-basis 

accidents would be derived from the 

unlikely event sequences from the PRA 

and then analyzed in a conservative 

approach by prescriptively assuming 

that only SR SSCs are available to 

mitigate postulated accident scenarios. 

Within the LMP methodology, event 

sequences with mean frequencies of 1 × 

10¥2/plant-year and greater would be 

classified as anticipated event 

sequences. Within the LMP 

methodology, infrequent event 

sequences with mean frequencies of 1 × 

10¥4/plant-year to 1 × 10¥2/plant-year 

would be classified as unlikely event 

sequences. ‘‘Very unlikely event 

sequences’’ would be less likely to occur 

than unlikely event sequences. Within 

the LMP methodology, rare event 

sequences with frequencies of 5 × 10¥7/ 

plant-year to 1 × 10¥4/plant-year would 

be classified as very unlikely event 

sequences. While the proposed 

terminology for these event sequences 

would create some differences between 

part 53 and the LMP, part 53 would use 

new terms for these event sequences 

specifically to avoid conflicts with 

terms already used within part 50 and 

part 52 to represent different concepts. 

Further, because some stakeholder 

comments demonstrated confusion 

related to the history of beyond-designbasis accidents terminology, these 

definitions seek to clarify the event 

categories in part 53. The sections of 

this preamble related to subparts B and 

C provide additional discussion of 

LBEs. 

B. Other General Provisions 

Section 53.040 would govern written 

communications and how applications 

and other required information must be 

submitted to the NRC. These 

requirements would be equivalent to 

those in § 50.4. 

Section 53.050 would establish 

requirements for enforcement action to 

which a licensee, an applicant, or a 

licensee’s or applicant’s contractor or 

subcontractor, or an employee of any of 

them may be subject for engaging in 

deliberate misconduct. These 

requirements would be equivalent to 

those in § 50.5. 

Section 53.060 would prohibit 

discrimination against an employee of a 

holder or applicant for an NRC license, 

permit, design certification (DC), or 

design approval, or a contractor or 

subcontractor of a holder or applicant 

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 §§ 50.7 and 52.5. 

Section 53.070 would govern the 

completeness and accuracy of 

information provided to the NRC. These 

requirements would be equivalent to 

those in §§ 50.9 and 52.6. 

Section 53.080 would govern 

exemptions from the requirements of 

the regulations in part 53. These 

requirements would be equivalent to 

those in §§ 50.12 and 52.7. 

Paragraphs (a) through (d) of § 50.90 

would establish requirements for 

standards that the NRC would consider 

in determining whether a construction 

permit (CP), operating license (OL), 

early site permit (ESP), combined 

license, or manufacturing license (ML) 

under part 53 would be issued to an 

applicant. These requirements would be 

equivalent to those in §§ 50.40, 50.42, 

50.43 and 50.22, respectively. 

Requirements equivalent to those in 

§§ 50.41 and 50.21 would not be 

included in part 53 because they apply 

to Class 104 licenses, and part 53 would 

not apply to those licenses. 

Section 53.100 would require that no 

license issued under part 53 would 

cover activities which are not under or 

within the jurisdiction of the United 

States. These requirements would be 

equivalent to those in § 50.53. 

Section 53.110 would state that 

licensees and applicants would not be 

required to provide design features or 

other measures for the specific purpose 

of protection against the effects of 

attacks and destructive acts by enemies 

of the United States directed against the 

facility or deployment of weapons 

incident to U.S. defense activities. 

These requirements would be 

equivalent to those in § 50.13. 

Section 53.115 would establish 

requirements for rights related to SNM. 

These requirements would be 

equivalent to those in § 50.54(b) and (c). 

Section 53.117 would establish 

requirements for license suspension and 

rights of recapture of the material or 

control of the facility in a state of war 

or national emergency declared by 

Congress. These requirements would be 

equivalent to those in § 50.54(d). 

Section 53.120 would establish 

requirements for information collection 

requirements and OMB approval. These 

requirements would be equivalent to 

those in § 50.8. 

Subpart B—Technology-Inclusive Safety 

Requirements 

Proposed subpart B, ‘‘TechnologyInclusive Safety Requirements,’’ would 

provide technology-inclusive safety 

criteria that would serve as performance 

standards for the subsequent 

performance-based requirements used 

throughout part 53. Subsequent subparts 

would define how specific activities 

during various stages of the life cycle of 

a commercial nuclear plant contribute 

to satisfying these high-level 

performance standards. The 

performance standards in subpart B 

would also establish a means to 

determine appropriate regulatory 

controls for SSCs, human actions, and 

programs in the following subparts. For 

example, the classification of SR SSCs 

would be built upon the proposed safety 

criteria in § 53.210, ‘‘Safety criteria for 

design-basis accidents.’’ The more 

detailed requirements for those SSCs 

would then be further defined in the 

design and analysis requirements in 

subpart C, ‘‘Design and Analysis 

Requirements.’’ The activities for 

manufacturing, constructing, and 

maintaining the SR SSCs would be 

governed by subpart E, ‘‘Construction 

and Manufacturing Requirements,’’ and 

subpart F, ‘‘Requirements for 

Operation.’’ 

Requirements for NSRSS SSCs 

warranting special treatment would 

likewise be determined under § 53.220, 

‘‘Safety criteria for licensing-basis 

events other than design-basis 

accidents,’’ in subpart B and § 53.460, 

‘‘Safety categorization and special 

treatment,’’ in subpart C. Regulatory 

requirements related to the NSRSS SSCs 

would be distinguished from the 

regulatory requirements for SR SSCs 

throughout part 53. Part 53 would afford 

more flexibility to applicants and 

licensees regarding how NSRSS SSCs 

would be used in the design and 

maintained during plant operations, as 

compared to SR SSCs. 

The collective set of performancebased requirements in part 53 would be 

sufficient, if met, for the NRC to make 

the findings required to grant an 

application for a utilization facility 

under Section 182 of the Act that the 

utilization of SNM will be in accord 

with the common defense and security 

and will provide adequate protection to 

the health and safety of the public. This 

construct would be similar to existing 

NRC regulations, which the Commission 

has said on many occasions do not 

specifically define ‘‘adequate 

protection.’’ However, compliance with 

NRC regulations may be presumed to 

assure adequate protection at a 


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minimum. The requirements throughout 

part 53 that support demonstrating 

compliance with § 53.220 would be 

similar to current regulations that both 

contribute to assuring adequate 

protection of public health and safety 

and are desirable to promote the 

common defense and security or to 

protect health or to minimize danger to 

life or property under Section 161 of the 

Act. 

Consistent with historical practice, 

Sections 182 and 161 of the Act are 

cited as authorizing legislation within 

this proposed rule. However, specific 

language from the Act would not be 

incorporated into the safety objectives 

or safety criteria in part 53. This is 

because, again consistent with historical 

practice, the NRC would not be defining 

‘‘adequate protection’’ through the 

individual safety requirements in part 

53. Rather, part 53 would enable the 

NRC to make its required findings under 

the Act by providing sufficient 

performance standards, safety criteria, 

and related requirements on how 

applicants must demonstrate 

compliance with subpart B and other 

subparts. 

Section 53.210 would provide safety 

criteria for DBAs that would be required 

to be identified under § 53.240 and 

analyzed under § 53.450(f) in subpart C 

of part 53. Subsequent sections in part 

53 would require that the SSCs relied 

upon to demonstrate compliance with 

the criteria in § 53.210 be classified as 

SR. The use of SR SSCs and the 25 rem 

reference values for potential 

radiological consequences would align 

with traditional deterministic 

approaches for LWRs from §§ 50.34, 

52.79, and 100.11 for evaluating the 

effectiveness of plant design features 

with respect to postulated reactor 

accidents. A footnote similar to that 

included in § 50.34(a)(1)(ii)(D)(1) and 

§ 52.79(a)(1)(vi)(A) would be included 

in § 53.210 to explain that the use of the 

25 rem value would not be intended to 

imply that this number constitutes an 

acceptable limit for an emergency dose 

to the public under accident conditions. 

Rather, this dose value has been set 

forth in this proposed section as a 

reference value that would be used in 

the evaluation of plant design features 

with respect to DBAs to verify that the 

proposed designs would provide 

assurance of low risk of public exposure 

to radiation in the event of an accident. 

The inclusion of the safety criteria for 

DBAs in subpart B would provide a 

logical structure supporting the 

identification and treatment of SR SSCs 

and establishing the corresponding 

functional design criteria for those 

SSCs. 

Section 53.220 would provide safety 

criteria for LBEs other than DBAs that 

would be required to be identified 

under § 53.240 and analyzed under 

§ 53.450(e) in subpart C. Whereas 

§ 53.210 and the related requirements 

for SR SSCs would provide that a 

defined success path exists for DBAs, 

the safety criteria for LBEs other than 

DBAs would establish the connections 

between SSC design, human actions, 

and programmatic controls and a 

broader set of potential internal and 

external hazards. These safety criteria 

would also address defense-in-depth 

matters such as a balanced 

consideration of prevention and 

mitigation. 

The safety criterion in § 53.220(b) 

would include a requirement to use a 

comprehensive risk metric or set of 

metrics and associated risk performance 

objectives against which calculated 

values of the risk metrics are compared. 

The comprehensive risk metrics or set of 

metrics and associated risk performance 

objectives would support a 

performance-based approach to 

developing an appropriate combination 

of design features and programmatic 

controls to prevent or mitigate LBEs 

other than DBAs. The applicant must 

propose the comprehensive risk metric 

or set of metrics and associated risk 

performance objectives, and the 

comprehensive risk metric or set of 

metrics and associated risk performance 

objectives must provide an appropriate 

level of safety. Comprehensive risk 

metrics should consist of a proposed 

plant risk metric or set of proposed risk 

metrics that approximate the total, 

overall risk from the facility and that 

address the range of possible plant 

configurations and associated internal 

and external hazards to the extent 

practicable. The associated risk 

performance objectives are 

preestablished, indicative values of the 

comprehensive risk metrics that are 

used as part of risk-informed decisionmaking. The methodology for 

developing and using proposed 

comprehensive risk metrics and 

associated risk performance objectives is 

defined by the proposed requirements 

for analyses in § 53.450. Therefore, the 

application must include a description 

of that methodology and, among other 

things, should explain the initial 

conditions, boundary conditions, and 

key assumptions used to develop and 

calculate the risk metrics. Screening 

tools and bounding or simplified 

methods may be used for any mode or 

hazard, provided that the applicant 

provides an acceptable technical basis. 

As with all risk-informed 

methodologies, treatment of 

uncertainties must be addressed. 

The risk performance objectives 

established under this methodology are 

likely to involve assessing and averaging 

the risks over a period of time (e.g., 

plant year) and would not constitute a 

real-time requirement that must be 

continuously demonstrated by the 

licensee. The use of a comprehensive 

risk metric or set of risk metrics and risk 

performance objectives that reflect an 

average risk to establish performance 

goals for SR and NSRSS SSCs is 

consistent with current practices that 

use other risk assessment techniques to 

address short-term plant configurations 

during plant maintenance activities. 

It is worth noting that the evaluation 

of plant risks, as represented by a 

comparison of analysis results to 

acceptable risk performance objectives 

for comprehensive risk metrics, would 

be one of several performance standards 

used in subpart B. The proposed use of 

multiple performance standards, 

including deterministic criteria and 

defense-in-depth measures, reflects an 

integrated decision-making process 

similar to that described in RG 1.174, 

‘‘An Approach for Using Probabilistic 

Risk Assessment in Risk-Informed 

Decisions on Plant-Specific Changes to 

the Licensing Basis,’’ Revision 3. The 

NRC’s approval of using a 

comprehensive risk metric or set of 

metrics with associated risk 

performance objectives is not, by itself, 

an indicator of adequate protection. 

Rather, the comparison of 

comprehensive risk metrics to 

associated risk performance objectives 

that are acceptable to the NRC is part of 

a suite of regulatory requirements that, 

when considered holistically, form the 

basis for the NRC’s decision-making. 

This is analogous to the approach used 

for plants licensed under part 50 and 

part 52, where no single regulatory 

requirement governs whether a plant is 

‘‘safe enough.’’ 

The RG 1.233, ‘‘Guidance for a 

Technology-Inclusive, Risk-Informed, 

and Performance-Based Methodology to 

Inform the Licensing Basis and Content 

of Applications for Licenses, 

Certifications, and Approvals for NonLight-Water Reactors,’’ describes an 

example of an acceptable approach for 

identifying and analyzing LBEs under 

part 50 and part 52, including the use 

of the quantitative health objectives 

(QHOs) stated in the NRC’s policy 

statement, ‘‘Safety Goals for Nuclear 

Power Plant Operation,’’ dated August 

4, 1986 (51 FR 28044), as corrected and 

republished August 21, 1986 (51 FR 

30028) (Safety Goals Policy Statement), 

as acceptable performance objectives for 


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comprehensive risk metrics. The use of 

comprehensive risk metrics, such as the 

individual early fatality risk (IEFR) and 

the individual latent cancer fatality risk 

(ILCFR), and associated risk 

performance objectives, such as the 

QHOs, from the Safety Goals Policy 

Statement, could form the basis for one 

approach to meet § 53.220(b). The 

requirement for comprehensive risk 

metrics, in combination with the other 

proposed requirements in subparts B 

and C, would bring the approach 

endorsed in RG 1.233 for parts 50 and 

52 into part 53. Additionally, the use of 

comprehensive risk metrics and 

associated risk performance objectives 

would provide a logical performance 

objective to support the risk 

management approaches in the various 

subparts comprising proposed part 53. 

The Commission stated in the 

introduction of the Safety Goals Policy 

Statement that improvements to thencurrent regulatory practices could lead 

to a more coherent and consistent 

regulation of nuclear power plants, a 

more predictable regulatory process, a 

better public understanding of the 

regulatory criteria that the NRC applies, 

and public confidence in the safety of 

operating plants. Accordingly, the 

Commission announced the safety goals 

with a focus on the risks to the public 

from nuclear power plant operation. 

Following the issuance of the Safety 

Goals Policy Statement, the NRC has 

used the comprehensive risk metrics 

and performance objectives provided in 

the safety goals within the criteria for 

many decisions involving safety 

judgments during the licensing and 

regulation of operating reactors and 

proposed nuclear reactor designs. 

Consistent with NUREG–0880, the 

proposed comprehensive risk metrics 

and associated risk performance 

objectives required under § 53.220(b) 

could be expressed in terms of a 

biologically average individual in terms 

of age and other risk factors. Although 

some comprehensive risk objectives 

such as the IEFR and ILCFR are defined 

in terms of fatality risks, the 

Commission continues to make clear 

that no death attributable to nuclear 

power plant operation will ever be 

‘‘acceptable’’ in the sense that the 

Commission would regard it as a routine 

or permissible event. Comprehensive 

risk metrics and associated risk 

performance objectives as used in this 

proposed rule would establish 

acceptable risks, not acceptable deaths. 

Applicants under the proposed part 

53 may choose to develop and seek NRC 

approval of comprehensive risk metrics 

or sets of risk metrics and associated 

risk performance objectives beyond 

those discussed above, including the 

use of surrogate measures for use in 

specific analyses to satisfy the proposed 

requirements in § 53.220(b). Such 

surrogate measures for comprehensive 

risk metrics and associated risk 

performance objectives could be used in 

a manner similar to the use of core 

damage frequency and conditional 

containment failure probability for 

LWRs within the safety goal evaluation 

process in NUREG/BR–0058, 

‘‘Regulatory Analysis Guidelines of the 

U.S. Nuclear Regulatory Commission,’’ 

and other assessments of LWRs using 

the NRC’s safety goals. The NRC would, 

as appropriate, review novel approaches 

for comprehensive metrics and 

associated risk performance goals 

proposed by applicants, industry 

organizations, or standard development 

organizations and would engage 

stakeholders during the development of 

the related regulatory guidance or 

specific licensing actions. 

Section 53.230 would require safety 

functions needed to ensure that the 

safety criteria under §§ 53.210 and 

53.220 can be met if an assumed LBE 

were to occur at a commercial nuclear 

plant. Section 53.230 would specify that 

limiting the release of radioactive 

materials from the facility is the primary 

safety function, and therefore, limiting 

potential offsite consequences (i.e., dose 

to a hypothetical individual) would be 

used as the primary performance metric 

throughout part 53. The additional or 

subsidiary safety functions needed to 

limit the release of radionuclides may 

include, without limitation, controlling 

processes related to reactivity, heat 

generation, heat removal, and chemical 

interactions. This proposed rule 

provides flexibility to applicants and 

licensees in identifying, implementing, 

and maintaining the safety functions 

supporting retention of radionuclides 

for commercial nuclear plants of varying 

sizes and technologies. 

Proposed § 53.240 would require 

applicants to identify and address LBEs. 

LBEs are unplanned events, resulting 

from both internal and external hazards, 

that are used in the design and analyses 

required under part 53 for licensing 

commercial nuclear plants. This ensures 

estimates of offsite consequences from 

analyses performed under proposed 

§ 53.450 are below the safety criteria 

identified under proposed §§ 53.210 and 

53.220 and that SSCs, personnel, and 

programs address the safety functions 

from proposed § 53.230. Including a 

high-level performance requirement 

related to the identification and analysis 

of LBEs in subpart B would reflect the 

historical and continuing importance of 

evaluating unplanned events as part of 

the licensing of commercial nuclear 

plants. Proposed § 53.240 would require 

identification and analysis of LBEs 

under § 53.450, which would require a 

PRA. Examples of acceptable methods 

of using PRAs to identify and assess 

LBEs would be the methodology in RG 

1.233, as discussed in Draft Regulatory 

Guide (DG)–1413, ‘‘TechnologyInclusive Identification of Licensing 

Events for Commercial Nuclear Plants.’’ 

Section 53.250 would establish 

defense-in-depth requirements based on 

the longstanding philosophy of 

providing defense in depth to address 

uncertainties about the design, 

operation, and performance of 

commercial nuclear plants. For 

example, parts 50 and 52 address 

defense in depth through layered 

prescriptive technical requirements 

(e.g., fuel performance, cladding 

integrity, reactor coolant system 

integrity, containment performance) for 

LWRs. In contrast, the flexibility 

afforded to applicants in how they 

propose to demonstrate compliance 

with the high-level safety criteria within 

part 53 would necessitate this specific 

requirement to ensure defense in depth 

is provided. The requirements in this 

section would state that no single 

engineered design feature, human 

action, or programmatic control, no 

matter how robust, should be 

exclusively relied upon to address LBEs 

other than DBAs. The phrase 

‘‘engineered design feature’’ would not 

preclude the possible crediting of 

inherent characteristics within the 

design and analysis for commercial 

nuclear reactors. While defense in depth 

would only be assessed for LBEs other 

than DBAs, the need to ensure 

dedicated success paths for DBAs would 

contribute to the overall defense in 

depth for each commercial nuclear plant 

under part 53. 

Section 53.260 would govern normal 

operations and would establish a level 

of safety based on current requirements 

in 10 CFR part 20, ‘‘Standards for 

Protection Against Radiation,’’ which 

limits doses to members of the public 

and dose rates in unrestricted areas. 

Section 53.270 would provide for the 

protection of plant workers and would 

establish a level of safety based on 

current requirements in 10 CFR part 20 

which limits occupational dose. 

Subpart C—Design and Analysis 

Requirements 

This subpart would provide 

requirements for the design of 

commercial nuclear plants and the 

supporting analyses, including the 

analyses of LBEs, to demonstrate that 

the performance standards in proposed 


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subpart B can be satisfied. The sections 

within subpart C would reflect the 

overall hierarchy throughout part 53, 

which would cover: (1) plant-level 

safety criteria (§§ 53.210, 53.220, and 

53.470); (2) safety functions (§ 53.230) 

needed to demonstrate compliance with 

the safety criteria; (3) design features 

(§ 53.400), human actions, and 

programmatic controls needed to fulfill 

the safety functions; and (4) functional 

design criteria (§§ 53.410 and 53.420) 

that must be defined for each design 

feature relied on to demonstrate the 

safety criteria (§§ 53.210, 53.220, and 

53.470) are met. Subpart C would also 

contribute to the logic and structure of 

part 53 by distinguishing between SR 

SSCs and NSRSS SSCs and licenseecontrolled programs that address LBEs 

other than DBAs. Specifically, SR SSCs, 

human actions, and programmatic 

controls needed to protect against DBAs 

are used to satisfy the safety criteria in 

§ 53.210. Non-safety-related but safetysignificant SSCs, human actions, and 

licensee-controlled programs that 

address LBEs other than DBAs generally 

contribute to the appropriate measures 

considering potential risks to public 

health and safety. 

Section 53.400 would establish a 

requirement that design features be 

provided for each commercial nuclear 

plant to satisfy the safety criteria and 

fulfill safety functions from proposed 

subpart B during LBEs. Other sections 

in subpart C would, in turn, further 

address the necessary capabilities and 

reliabilities for SSCs by establishing 

functional design criteria, fulfilling 

design requirements, performing 

analyses of LBEs, performing other 

supporting analyses, and categorizing 

SSCs based on their roles in preventing 

or mitigating LBEs. 

Section 53.410 would require that 

functional design criteria be defined for 

design features relied upon to 

demonstrate that the consequences from 

DBAs would be below the criteria in 

§ 53.210 through analyses performed 

under § 53.450(f), which includes 

insights from both PRAs and 

deterministic analyses. Other sections 

within part 53 would establish 

appropriate controls on these design 

features (e.g., safety classification, 

protection from external hazards, 

quality assurance, and TS) to ensure the 

functional design criteria are satisfied. 

The performance requirements for the 

SSCs needed to address DBAs and the 

corresponding human actions and 

programmatic controls would contribute 

to ensuring that a commercial nuclear 

plant licensed under part 53 would 

meet the safety criteria in § 53.210. 

Section 53.415 would require that SR 

SSCs be protected against or designed to 

withstand the effects of natural 

phenomena (e.g., earthquakes, 

tornadoes, hurricanes, floods, tsunami, 

and seiches) and constructed hazards 

(e.g., from dams, transportation routes, 

and military or industrial facilities). 

Specifically, § 53.415 would require that 

SR SSCs remain capable of performing 

the safety functions stated in § 53.230 

for which they are credited up to the 

design-basis external hazard levels as 

determined under § 53.510. As used in 

§ 53.415 and subpart D of part 53, a 

hazard level would refer to such things 

as the magnitude and recurrence rate of 

an earthquake and the resultant ground 

motions, the height of a flood, the force 

of hurricane winds, or the 

concentrations of chemicals resulting 

from a release from a nearby facility. 

These requirements would support 

either traditional deterministic 

approaches for determining and 

protecting against external hazards or 

probabilistic approaches that are being 

developed for seismic and some other 

external hazards. 

Section 53.420 would require that 

functional design criteria be defined for 

design features that play a significant 

role in demonstrating that the safety 

criteria for LBEs other than DBAs are 

satisfied. The analyses required for this 

demonstration would be described in 

proposed § 53.450(e), which would 

require that those events be identified 

and assessed using a PRA methodology 

in combination with other generally 

accepted approaches for systematically 

evaluating engineered systems. The 

SSCs determined to be safety significant 

(i.e., either SR or NSRSS) would have 

associated special treatment 

requirements as specified in § 53.460. 

Special treatment would be defined in 

subpart A of part 53 and generally refers 

to measures (e.g., quality assurance, 

testing, monitoring) taken beyond the 

procurement and installation of 

commercial grade products to provide 

confidence that the SSC will comply 

with the applicable functional design 

criteria. The inclusion of a systematic 

approach to identifying the functional 

design criteria for SSCs and tailoring the 

special treatments to specific LBEs and 

safety functions is an important 

contributor to satisfy the proposed 

safety criteria in subpart B. Therefore, 

designers and licensees for commercial 

nuclear plants would be provided 

flexibility on how LBEs other than 

DBAs are either prevented or mitigated 

and how the calculated comprehensive 

plant risks satisfy the safety criterion 

established under § 53.220(b). 

Section 53.425 would establish 

requirements for design features and 

related functional design criteria 

limiting doses to members of the public 

during normal operations to satisfy the 

criteria in part 20. Section 53.430 would 

provide similar requirements for design 

features and related functional design 

criteria for protection of plant workers 

to meet the safety criteria in part 20. 

Similar to existing regulations, the NRC 

considers that licensees would generally 

comply with the requirements of part 20 

to keep doses as low as reasonably 

achievable by meeting a design objective 

of keeping doses to the public from 

routine plant effluents less than 10 

millirem per year. This goal is similar to 

that provided by appendix I to part 50 

and would assist designers, applicants, 

and licensees in performing the 

evaluations of possible reductions in 

public dose from routine effluents when 

considering costs and other factors. As 

emphasized in existing regulations in 

part 50, the design objective of keeping 

doses to the public from routine plant 

effluents less than 10 millirem per year 

should not be construed as a radiation 

protection standard. The NRC 

anticipates that future guidance will 

continue to reflect this performance 

goal. 

The proposed requirements in 

§§ 53.425 and 53.430 for design features 

and functional design criteria to support 

radiation protection activities have 

parallels in existing regulations such as 

§ 50.34(a) and (b)(3), which require in 

part that the means be provided for 

meeting the requirements of part 20 and 

General Design Criterion 60, 61, 63, and 

64 in appendix A to part 50, which 

provide radiation protection related 

design criteria. 

Section 53.440 would address various 

design requirements that warrant 

specific mention to ensure that the 

design features required by § 53.400 

comply with the functional design 

criteria required by §§ 53.410 and 

53.420. These requirements would be 

met through design practices, 

consideration of testing and operating 

experience, and various assessments of 

LBEs and other potential challenges to 

commercial nuclear plants. Discussions 

of some of the key design requirements 

included in this section follow. 

• § 53.440(a): An essential element to 

ensuring a proposed design can comply 

with the performance criteria in 

proposed part 53 would be that the 

abilities of design features to fulfill their 

safety functions are demonstrated by a 

combination of analyses, test programs, 

prototype testing, and operating 

experience. This requirement closely 

aligns with the language in § 50.43(e) 


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and is proposed in part 53 as the same 

foundational requirement. In addition, 

the proposed § 53.440(a) would require 

the design processes for SSCs under this 

section to include administrative 

procedures for evaluating operating, 

design, and construction experience for 

considering applicable important 

industry experiences in the design of 

those SSCs. This proposed requirement 

corresponds to the existing requirement 

under § 50.34(f)(3)(i) that was developed 

in response to the 1979 accident at 

Three Mile Island Nuclear Generating 

Station. 

• § 53.440(b): The design and 

licensing of commercial nuclear plants 

should use generally accepted 

consensus codes and standards. Such 

codes and standards ensure sufficient 

testing and qualification of materials 

and equipment and provide defined 

processes, specifications, and 

acceptance criteria for use by designers 

and suppliers. The NRC would indicate 

acceptance of consensus codes and 

standards used in the design and 

licensing of a specific commercial 

nuclear plant either through the NRC’s 

generic endorsement of a code or 

standard (i.e., through regulatory 

guidance), including any limitations or 

conditions, that can be referenced 

within an application, or through the 

review of a referenced code or standard 

as part of the review of a specific 

application. 

• § 53.440(c): The design 

requirements in subpart C would 

require the materials used for SR and 

NSRSS SSCs to be qualified for their 

service conditions over the design life of 

the SSC. 

• § 53.440(d): The requirements in 

§ 53.440 would include the need to 

consider possible degradation 

mechanisms for materials and 

equipment to inform both the design 

process and the development of 

integrity assessment programs to be 

executed during plant operations in 

accordance with subpart F of part 53. 

The inclusion of requirements related to 

designing and monitoring for possible 

degradation mechanisms reflects 

important lessons learned from the 

history of LWRs as well as operating 

experience with structures and systems 

in countless other engineering 

endeavors. 

• § 53.440(e) and (f): The design 

requirements in subpart C would state 

specific design requirements similar to 

existing requirements in parts 50, 52, 

and 73 for protections against fires and 

explosions and consideration of safety 

and security together in the design 

process. 

• § 53.440(g) and (h): Specific design 

requirements are proposed to ensure 

that commercial nuclear reactors under 

part 53 have the capability to achieve 

and maintain subcriticality and longterm cooling. The requirements would 

be included to address the potential that 

some reactor designs may be able to 

achieve a stable end state for the 

purpose of event analyses but might 

need further actions to completely shut 

down and service the facility. 

• § 53.440(i): The design, analysis, 

and development of programmatic 

controls under part 53 would consider 

the number of reactor units and other 

significant inventories of radioactive 

materials contributing to the risks to 

public health and safety. This would 

reflect the definition of ‘‘Commercial 

nuclear plant’’ in subpart A and 

reinforce that the evaluation of LBEs is 

performed on a plant-wide basis. This 

aspect of part 53 would be different 

from parts 50 and 52, which generally 

define safety requirements on the 

assumption of events involving only 

individual reactor units. 

• § 53.440(j): A design requirement is 

proposed to provide a technologyinclusive requirement that would be 

equivalent to the requirements in 

§ 50.150 to address the possible impact 

of a large commercial aircraft. 

• § 53.440(k): The inclusion of a 

specific proposed requirement to 

address the risks to public health from 

potential chemical hazards of licensed 

material is appropriate given the 

diversity of reactor technologies and 

designs that might be licensed under 

part 53. The requirement in part 53 

would be similar to the existing 

requirements in 10 CFR part 70, 

‘‘Domestic Licensing of Special Nuclear 

Material,’’ that address both potential 

radiological and chemical hazards for 

licensed materials at fuel cycle facilities. 

• § 53.440(l): Provisions are proposed 

to require that measures be taken during 

the design of commercial nuclear plants 

to minimize contamination of the 

facility and the environment, facilitate 

eventual decommissioning, and 

minimize the generation of radioactive 

waste in accordance with § 20.1406. 

• § 53.440(m): A design requirement 

is proposed to provide a technologyinclusive equivalent to the requirements 

in § 50.68 by including options for 

commercial nuclear plants to either 

have a monitoring system capable of 

detecting a criticality as described in 

§ 70.24 or to have restrictions on SNM 

handling and storage that would prevent 

inadvertent criticality events. 

• § 53.440(n): The design would need 

to reflect state-of-the-art human factors 

principles for safe and reliable 

performance in all settings that human 

activities are expected for performing or 

supporting the continued availability of 

plant safety or emergency response 

functions. 

Section 53.450 would establish 

analysis requirements and would center 

upon the use of a PRA in combination 

with other generally accepted 

approaches for systematically evaluating 

engineered systems. The reliance on 

PRAs as a key component in the 

proposed analysis requirements for part 

53 would reflect the decades of 

improvements in PRA methodologies 

and the increasing use of PRA 

techniques in the design, licensing, and 

oversight of both operating and future 

nuclear reactors. Part of the 

Commission’s PRA Policy Statement is 

that the use of PRA technology should 

be increased in all regulatory matters to 

the extent supported by the state of the 

art in PRA methods and data and in a 

manner that complements the NRC’s 

deterministic approach and supports the 

NRC’s traditional defense-in-depth 

philosophy. The need to supplement 

PRA insights with other engineering 

approaches and judgments reflects the 

NRC’s longstanding policy described in 

the SRM to SECY–98–144, ‘‘Staff 

Requirements—SECY–98–144—White 

Paper on Risk-Informed and 

Performance-Based Regulations,’’ dated 

February 24, 1999, for regulatory 

decision-making to be risk-informed but 

not solely based on numerical results of 

a risk assessment (i.e., not a risk-based 

approach). Part 53 would maintain a 

role for NRC’s traditional deterministic 

approaches (particularly for DBAs) and 

defense-in-depth philosophy by 

including specific requirements 

utilizing these regulatory tools in 

subparts B and C. 

PRA would be used in combination 

with other techniques in part 53 to 

identify and categorize LBEs, classify 

SSCs, and evaluate defense in depth. 

This increased role for the PRA 

necessitates that it would be developed, 

performed, and maintained in 

accordance with NRC-approved 

standards and practices (see § 53.450(c) 

and (d)). The computer codes used to 

model the plant response and the 

behavior of the barriers to the release of 

radionuclides would need to be 

qualified for the range of conditions 

being simulated across a wide range of 

unplanned events. These analyses 

would need to use realistic approaches 

and address uncertainties associated 

with states of knowledge, modeling, and 

performance of SSCs. 

While industry consensus PRA 

standards and peer review processes 

endorsed in RGs 1.200 and 1.247 remain 


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acceptable for developing a PRA, they 

are not regulatory requirements and an 

application under part 53 need not 

follow every aspect of the applicable 

consensus PRA standard. Existing 

processes for defining the scope and 

capability of a PRA supporting an 

application offer flexibility in 

determining the degree to which the 

PRA needs to be developed and may be 

informed by other factors such as design 

complexity and the needed degree of 

realism and level of detail, consistent 

with the use of the PRA and substance 

of the application. Such processes are 

currently available for appropriately 

defining the scope of the PRA and 

determining applicability of supporting 

requirements in consensus PRA 

standards needed to satisfy the 

proposed regulatory requirements for 

the specific uses of analyses under 

§ 53.450(b). Likewise, NRC 

determinations of the acceptability of 

such PRAs would include consideration 

of the appropriateness of the applicantdefined scope as part of determining the 

applicability of and conformance to 

consensus PRA standard supporting 

requirements consistent with the 

current state of practice. In addition, 

these determinations would include 

consideration of other aspects of the 

development of the PRA, such as PRA 

peer reviews. An NRC determination of 

the acceptability of a PRA includes but 

is not limited to assessing the initial and 

boundary conditions and key 

assumptions used in the analysis, 

treatment of uncertainties, and the use 

of screening tools and bounding or 

simplified methods for any mode or 

hazard, provided the use of those tools 

and methods is justified by an 

acceptable technical basis. In that 

regard, the consensus PRA standards 

would not be applied by the NRC as a 

strict checklist of requirements for part 

53 PRA acceptability determinations. 

The proposed § 53.450(c) would 

require periodic maintenance and 

upgrading of the PRA to maintain an 

alignment between the supporting 

analyses and the design and 

performance of plant equipment, 

programs and procedures, and other 

factors associated with meeting the 

safety criteria of the proposed § 53.220 

and the evaluation criteria of proposed 

§ 53.450(e)(2). The periodic 

maintenance of the PRA would also be 

a means to consider new or revised 

information related to external hazards, 

industry operating experience, 

performance issues with or degradation 

of SSCs, and other contributors to the 

frequency and potential consequences 

of various event sequences. The 

periodic assessments performed by 

licensees to support the maintenance of 

the PRA and other requirements in the 

proposed part 53 would be 

complemented by NRC inspections and 

programs to assess new or revised 

information related to topics such as 

natural hazards, operating experience, 

and potential generic safety issues. 

The categories of LBEs used in part 53 

would include anticipated event 

sequences, unlikely event sequences, 

and very unlikely event sequences. The 

unlikely event sequences would include 

those events with estimated frequencies 

well below the frequency of events 

expected to occur during the lifetime of 

a commercial nuclear plant. An 

important aspect of the analysis 

requirements is that, under proposed 

§ 53.450(e), the analyses of LBEs other 

than DBAs would not only be used to 

show the performance criteria of 

§ 53.220 are satisfied but to also show 

that evaluation criteria defined for each 

LBE or category of LBEs would also be 

satisfied. Such evaluation criteria for 

specific LBEs or categories of LBEs 

would be defined in terms of limits on 

the release of radionuclides or 

maintaining the integrity of one or more 

barriers used to limit the release of 

radionuclides and reflect a graded 

approach of allowing lesser potential 

consequences from more frequent 

events. An example of such evaluation 

criteria for a range of LBEs that could 

likely be expanded for part 53 is 

provided in RG 1.233. Another 

proposed requirement for the proposed 

§ 53.450(e) analyses is that the 

methodology would need to include a 

means to identify event sequences 

deemed risk-significant such that those 

event sequences can be given special 

attention within other sections of part 

53. 

Part 53 would maintain an important 

role for a deterministic analysis of DBAs 

in the performance criteria of § 53.210 

and the related analytical requirements 

in § 53.450(f). The analysis of DBAs 

would be required to address event 

sequences drawn from those with 

estimated frequencies below the 

expected lifetime of a generation of 

reactors (e.g., event sequences with 

frequencies as low as one in ten 

thousand years). As proposed in this 

section, DBAs would need to be 

analyzed using deterministic methods 

and ensure a safe, stable end state with 

reliance upon only SR SSCs and human 

actions, if needed, to be performed by 

operators licensed under the provisions 

of §§ 53.760 through 53.795. 

While the DBAs analyzed under part 

53 would be similar to the traditional 

DBAs analyzed under parts 50 and 52, 

there are important distinctions between 

the overall role of DBA analyses in part 

50 and proposed part 53. In part 53, the 

role of the DBA analysis would be more 

narrowly focused on selecting SR SSCs 

and determining functional design 

criteria for those SSCs to ensure the 

commercial nuclear plant meets the 

safety criteria in § 53.210. The overall 

control of risks posed by commercial 

nuclear plants under part 53 would be 

provided by the analyses of and 

measures taken for both DBAs and other 

LBEs, including very unlikely event 

sequences. This would contrast with the 

traditional deterministic approach in 

part 50 wherein the analyses of DBEs 

such as DBAs were used to provide 

bounding assessments, incorporate 

standard design rules such as 

assumptions related to single failures, 

and to define conservative performance 

requirements for SR SSCs. Limitations 

related to the traditional deterministic 

approach were addressed in part 50 

through case-by-case assessments and 

specific actions for beyond-design-basis 

events such as anticipated transients 

without scram and station blackout. 

Section 53.450 would also include 

provisions to ensure that analyses are 

performed to support the design 

requirements of § 53.440(e) on fire 

protection, § 53.440(j) on aircraft impact 

assessments, and § 53.425 on using 

design features and plant programs to 

control doses to members of the public 

from routine effluents and direct 

radiation from contained sources. The 

proposed analysis requirements related 

to fire protection would support either 

a traditional, deterministic approach or 

a more risk-informed approach where 

the risks from fires are addressed within 

the identification and analyses of LBEs. 

Section 53.460 would establish 

criteria for the safety classification of 

SSCs and determination of appropriate 

special treatments. As noted in subpart 

A, the term ‘‘Special treatments’’ would 

be defined to mean those items, such as 

measures taken to satisfy functional 

design criteria, quality assurance, and 

programmatic controls, which provide 

assurance that certain SSCs will provide 

defense in depth or perform risksignificant functions. These 

requirements would also provide 

confidence that the SSCs will perform 

under the service conditions and with 

the reliability credited in the analysis 

performed in accordance with § 53.450 

to satisfy the safety criteria in §§ 53.210 

and 53.220. The terminology used in 

part 53 would include the following 

categories for SSC classification: (1) SR; 

(2) NSRSS; and (3) non-safety 

significant. Requirements for SR SSCs 

would be defined in other sections of 


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part 53 and would include using TSs for 

controls during operation and the 

application of quality assurance 

requirements from appendix B of part 

50. 

Requirements for NSRSS SSCs would 

include the need to identify necessary 

special treatments such as performance 

measures on reliability. Licensees 

would generally be afforded flexibility 

in maintaining and changing special 

treatments for SSCs categorized as 

NSRSS. Non-safety-significant SSCs 

would be addressed under normal 

licensee programs for commercial grade 

equipment and typical industry 

practices for general plant design and 

maintenance. Safety-related SSCs would 

also contribute to defense in depth and 

risk-significant functions and may 

warrant special treatments beyond those 

defined for their SR functions to reflect 

their role in meeting the safety criteria 

in § 53.220 and the evaluation criteria in 

§ 53.450(e). 

Section 53.470 would allow an 

applicant or licensee to seek operational 

flexibilities by adopting more restrictive 

criteria than those provided in § 53.220 

and that might otherwise be used in the 

analysis of LBEs under § 53.450(e). Such 

an approach might be taken to ensure 

sufficient safety margins to gain 

operational flexibilities in areas such as 

justifying siting in relation to 

population centers or staffing levels. As 

an example, an applicant or licensee 

could propose to justify siting proposals 

by adopting alternate criteria for very 

unlikely event sequences. Such 

alternate criteria could require 

calculated consequences for an 

individual at the exclusion area 

boundary to be less than one rem total 

effective dose equivalent (TEDE). This 

section would establish requirements to 

ensure that, if more restrictive 

evaluation criteria than those required 

by a methodology were used to justify 

operational flexibilities, then the 

analysis, design features, and 

programmatic controls would be 

established and maintained accordingly. 

Section 53.480 would establish 

seismic design considerations. This 

proposed section would relate to the 

safety criteria in subpart B, the 

analytical requirements related to 

external hazards in § 53.450, and 

subpart D, ‘‘Siting Requirements.’’ For 

licenses issued under part 53, this 

section in subpart C would support a 

variety of approaches to seismic design. 

For example, a design for a commercial 

nuclear plant could show that SSCs are 

able to withstand the effects of 

earthquakes by adopting an approach 

similar to that in appendix S to part 50. 

Alternatively, an applicant could follow 

the more recent risk-informed 

alternatives afforded by standards 

development organizations (e.g., 

American Society of Civil Engineers 

(ASCE)/Structural Engineering Institute 

(SEI) 43–19, ‘‘Seismic Design Criteria for 

Structures, Systems, and Components in 

Nuclear Facilities.’’) Because the agency 

has not endorsed ASCE/SEI–43–19, an 

applicant can propose to use ASCE/SEI 

43–19 on an application specific basis to 

meet § 53.480 and the NRC would 

evaluate the adequacy of the standard as 

applied in that application. The design 

could also be done with the full 

integration of seismic PRAs into the 

design and licensing of a particular 

commercial nuclear plant. This section 

has been developed to accommodate a 

variety of potential risk-informed, 

performance-based seismic design 

approaches. The analyses required by 

§ 53.450 would need to address seismic 

hazards as well as other external 

hazards. The expected responses of 

SSCs to a range of seismic events would 

be included in the analyses when 

ensuring that the safety criteria defined 

under § 53.220 would be met. The 

potential SSC responses to seismic 

hazards could be addressed in the 

analyses using a fragility model 

(conditional probability of its failure at 

a given hazard input level), a high 

confidence of low probability of failure 

value, or other method endorsed or 

otherwise found acceptable by the NRC. 

Subpart D—Siting Requirements 

Proposed subpart D in part 53 would 

state requirements for the siting of 

commercial nuclear plants and would 

serve the role provided by 10 CFR part 

100, ‘‘Reactor Site Criteria,’’ for nuclear 

reactors licensed under parts 50 and 52. 

As reflected in proposed § 53.500, the 

reason for establishing siting 

requirements would remain the same as 

it has been historically, which is to 

ensure that licensees and applicants 

assess what impact the site environs 

may have on a commercial nuclear plant 

(e.g., external hazards) and, conversely, 

what potential adverse health and safety 

impacts a commercial nuclear plant may 

have on nearby populations in view of 

the site characteristics. 

Proposed § 53.510 would require that 

design-basis external hazard levels be 

identified and characterized based on 

site-specific assessments of natural and 

constructed hazards with the potential 

to adversely affect plant functions. The 

site-specific assessments would be used 

in the proposed § 53.415, which would 

require that SR SSCs be designed to 

withstand the effects of natural 

phenomena and constructed hazards of 

levels or severities up to design-basis 

external hazard levels. The design-basis 

levels for external hazards relevant to a 

site would need to account for 

uncertainties and variabilities in data, 

models, and methods used to 

characterize those hazards. Existing 

approaches could be used to 

demonstrate compliance with this 

requirement. The historical importance 

of assessing seismic events as risks to 

commercial nuclear plants and the 

associated development of riskinformed approaches to address seismic 

events would be reflected in proposed 

§ 53.480, ‘‘Earthquake engineering,’’ and 

specific requirements in subpart C. The 

NRC is developing a graded approach 

for seismic design by grouping SSCs 

into different seismic design categories 

(SDCs) based on their risk significance. 

While the agency has not endorsed 

ASCE/SEI–43–19, an applicant can 

propose to use ASCE/SEI 43–19 on an 

application-specific basis to meet 

§ 53.480 and the NRC will evaluate the 

adequacy of the standard as applied in 

that application. The NRC staff will 

continue to review ASCE/SEI–43–19 as 

part of its efforts to further develop 

guidance in this area. The approach 

described in RG 1.208, ‘‘A PerformanceBased Approach to Define the SiteSpecific Earthquake Ground Motion,’’ 

would be an acceptable way to develop 

site-specific ground motion response 

spectra for SSCs under appendix S to 

part 50, which corresponds to SSCs that 

are categorized as the highest SDC 

(SDC–5) in ASCE/SEI 43–19. 

The evaluation of seismic hazards 

under subpart D would need to be 

sufficient to inform a site-specific 

design (e.g., a CP or custom COL) or 

confirm the use of a standard design for 

a commercial nuclear plant under 

§ 53.480 and other sections of subpart C. 

A risk-informed approach could use 

several design-basis ground motions 

(DBGMs) to assess SSCs in various SDCs 

(i.e., one DBGM per SDC). Section 

53.510(d) would state that geologic and 

seismic siting factors must also include 

related hazards such as seismically 

induced flooding and volcanic activity 

that may affect the design and operation 

of a proposed commercial nuclear plant 

for the proposed site. 

Section 53.520 would require 

applicants to identify and assess site 

characteristics related to topics which 

might include meteorology, geology, 

hydrology, or other areas in the design 

and analyses required under subpart C. 

Proposed section 53.530 would set 

requirements for population-related 

considerations and maintain 

requirements and definitions similar to 

those currently in part 100 for an 

exclusion area, low population zone, 


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1On 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. 

and population center distance. The 

NRC recognizes that some applicants 

may propose to essentially collapse the 

exclusion area and low population zone 

to the site boundary. This approach 

would rest on a demonstration that the 

calculated consequences of DBAs 

remain below the proposed dose 

guidelines used in § 53.210, which are 

the same as those in the existing 

regulations in parts 50, 52, and 100. The 

proposed definitions in § 53.020 would 

allow such configurations, assuming 

they were justified by the design and 

analyses from subpart C. This approach 

should provide flexibility to justify 

alternative exclusion areas and low 

population zones without foreclosing 

the option for an applicant to define 

more conventional exclusion areas and 

low population zones outside of a 

defined site boundary. The NRC’s longstanding preference for siting reactors in 

areas of low population density would 

be maintained in part 53 by using the 

current language from part 100 in 

proposed § 53.530(c). The NRC revised 

guidance related to population densities 

surrounding a commercial nuclear plant 

in Revision 4 to RG 4.7, ‘‘General Site 

Suitability Criteria for Nuclear Power 

Stations’’ to reflect Commission 

direction in SRM–SECY–20–0045, 

‘‘Population Related Siting 

Considerations for Advanced Reactors.’’ 

Site-related requirements in part 20 

(restricted area) and part 73 (protected 

and owner-controlled areas) would 

remain applicable to commercial 

nuclear plants licensed under part 53. 

Proposed section 53.540 would 

require that site characteristics be 

appropriately considered in other 

activities such as the design and 

analysis performed under proposed 

subpart D and the emergency planning 

and security programs under proposed 

subpart F. 

Subpart E—Construction and 

Manufacturing Requirements 

The proposed part 53 language would 

establish construction and 

manufacturing requirements in subpart 

E. The proposed language for 

construction-related activities would 

largely reflect current requirements in 

part 50 without any fundamental 

changes. Limited changes would be 

made in several places, as described in 

the following paragraphs, to be 

technology-neutral and for consistency 

with the organization and language of 

part 53. The proposed language for 

requirements for manufacturing 

activities would largely mirror those for 

construction-related activities. However, 

the proposed manufacturing 

requirements have been updated from 

the current requirements in subpart F of 

part 52 to better accommodate the 

possible factory fabrication of 

manufactured reactors. The 

manufacturing of specific components 

outside the scope of an ML would not 

be addressed by these proposed 

subparts. 

Section 53.600 would establish the 

overall construction and manufacturing 

requirements for CPs, OLs, COLs, MLs, 

and limited work authorizations 

(LWAs). This section would connect the 

construction and manufacturing 

requirements to the safety criteria, 

quality assurance requirements, and 

other requirements located in other 

subparts. These requirements would 

require that construction and 

manufacturing activities be managed 

and conducted such that when 

combined with associated design 

features and programmatic controls, the 

constructed plant would satisfy the 

relevant requirements in subpart B. 

Section 53.605 would establish 

requirements for the reporting of defects 

and instances of noncompliance during 

construction. This section would 

provide equivalent requirements to 

those in § 50.55(e). 

Section 53.610(a) would establish the 

requirement to have in place a welldefined command and control structure 

to manage construction activities. The 

requirements would generally reflect 

current requirements, with an emphasis 

on the quality assurance programs for 

complying with the requirements in 

appendix B to part 50. The proposed 

§ 53.610(a)(6) would require 

programmatic controls for implementing 

special treatment for NSRSS SSCs to 

align with requirements in other 

subparts in part 53. The section would 

also refer to other NRC regulations to 

address matters such as requirements to 

have a FFD program, a radiation 

protection program if radioactive 

materials are brought onto the site, and 

security programs to protect sensitive 

information and protect against cyber 

threats. 

Section 53.610(b) would provide 

requirements governing construction 

activities, including the equivalent of 

the requirement in § 50.10(e) that 

prohibits starting construction until the 

NRC has authorized the activities by 

issuing a CP, COL, ESP, or LWA. 

Section 53.610(b)(1)(iii) would require 

procedures to be in place prior to 

beginning construction to ensure that 

construction-related activities do not 

undermine important features such as 

slope stability and that constructionrelated activities such as backfilling of 

excavated portions of the site 

appropriately address potential preconstruction activities such as the 

emplacement of retaining walls or 

drainage systems. Other requirements in 

these paragraphs would be equivalent to 

requirements in parts 50 and 52 with 

appropriate references to other parts for 

items such as possession of byproduct 

material or SNM, protecting operating 

units from construction activities for 

commercial nuclear plants with 

multiple reactor units, and having a 

redress plan in case LWA activities are 

terminated. 

Section 53.610(c) would address 

inspection and acceptance activities by 

including requirements in part 53 

equivalent to specific quality assurance 

criteria in appendix B to part 50 and 

inspections, tests, analyses, and 

acceptance criteria (ITAAC) in part 52 

for COLs. 

Section 53.620(a) would include 

proposed requirements covering the 

activities performed under an ML issued 

under part 53. Provisions related to MLs 

were first adopted by the NRC in 1973 

through the addition of appendix M to 

part 50. The regulation supported the 

manufacture of a nuclear power reactor 

to be incorporated into a commercial 

nuclear plant under a CP and operated 

under an OL at a different location from 

the place of manufacture.1 The 

regulations and processes for MLs were 

changed substantially in the part 52 

rulemaking in 2007 (72 FR 49352). The 

most important shift in the ML concept 

in that rulemaking was that a final 

reactor design, which would be 

equivalent to that required for a 

standard DC under part 52 or an OL 

under part 50, must be submitted and 

approved before issuance of an ML. The 

rationale for that change was that 

approval of a final design ensures early 

consideration and resolution of 

technical matters before there is any 

substantial commitment of resources 

associated with the actual manufacture 

of the reactor, which greatly enhances 

regulatory stability and predictability. 

The proposed part 53 sections in 

subpart E for manufacturing and in 

subpart H for licensing matters would 

maintain requirements equivalent to 

those in part 52 for MLs. The NRC 

approval of a standard design and 

related manufacturing processes, 

coupled with a stable workforce and 

established procedures, has the 

potential for maintaining and even 

improving the quality and consistency 

of manufacturing, as compared to the 

traditional method of constructing 


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reactors onsite by a variety of 

contractors and subcontractors. 

Subpart E would include 

requirements that would apply to 

portions of a manufactured reactor in 

recognition that some activities covered 

by an ML may occur at different 

fabrication facilities. As with the 

preceding sections on construction, 

§ 53.620 would establish the 

requirements to have in place programs, 

procedures, and a well-defined 

command and control structure to 

manage manufacturing-related 

activities. 

Section 53.620(b) in subpart E would 

propose requirements for executing the 

manufacturing activities following 

receipt of an ML under part 53. 

Information about the design and 

manufacturing processes should be 

provided by the applicant. The 

importance of the ML is reflected in 

several of the proposed requirements in 

§ 53.620(b) that would refer to 

complying with the ML, including 

conducting manufacturing processes 

within facilities for which the license 

holder can control activities. The 

essential role of post-manufacturing 

inspections would also be incorporated 

into this proposed section by requiring 

the holder of the ML to perform 

inspections and have acceptance 

processes for manufactured reactors or 

portions of a manufactured reactor. 

Section 53.620(c) would provide 

proposed requirements for the control of 

radioactive materials if the holder of an 

ML plans to possess and use source, 

byproduct, or SNM as part of the 

manufacturing process. By and large, 

the proposed subpart E would refer to 

NRC regulations in 10 CFR part 30, 

‘‘Rules of General Applicability to 

Domestic Licensing of Byproduct 

Material,’’ 10 CFR part 40, ‘‘Domestic 

Licensing of Source Material,’’ and part 

70 for the requirements on controlling 

radioactive materials. Several specific 

requirements to address the potential 

hazards of radioactive materials are 

proposed in areas such as having a fire 

protection program, an emergency plan, 

training programs, and procedures to 

minimize contamination. 

The most significant change proposed 

for MLs in part 53 as compared to MLs 

under part 52 relates to § 53.620(d) in 

subpart E and the associated licensing 

provisions in subpart H. These 

provisions would allow and establish 

requirements for the loading of fuel into 

a manufactured reactor at the 

manufacturing site for subsequent 

transport to a commercial nuclear 

facility that will operate pursuant to a 

COL. The first requirement in the 

proposed § 53.620(d) would establish 

limitations on when a license under part 

70 would authorize the loading of fuel 

into a reactor manufactured under an 

ML. The proposed regulation would 

require the manufactured reactor to 

include at least two independent 

physical mechanisms that will each 

prevent criticality should conditions 

most favorable to critical operation be 

introduced (e.g., optimum neutron 

moderation and reflection). This 

requirement would contribute to the 

NRC’s longstanding practice of requiring 

defense in depth for preventing 

accidents in any facility dealing with 

SNM, including requirements in § 70.64 

for certain part 70 licensees to adhere to 

the ‘‘double contingency principle.’’ 

The requirements to have in place 

mechanisms to prevent criticality could 

likewise support meeting other 

provisions in subpart H to part 70, such 

as those related to having a safety 

program and integrated safety 

assessment. The mechanisms to 

preclude criticality in the proposed 

requirements would reasonably ensure 

that a manufactured reactor would not 

become critical assuming optimum 

neutron moderation, and optimum 

neutron reflection conditions. With the 

proposed requirements for mechanisms 

to prevent criticality and all criticality 

safety controls required by 10 CFR part 

70 in place, the presence of fuel in the 

manufactured reactor would not create 

a nuclear hazard different than the 

hazard from the presence of the same 

fuel in a storage location or container 

licensed under 10 CFR part 70. 

Collectively, the proposed measures 

would reasonably ensure that the 

manufactured reactor would not be 

capable of operations, thereby obviating 

the need for a COL under §§ 53.1416 

and 53.1440 to authorize fuel loading. 

Additionally, this approach would focus 

the ML application and its review on 

the design, manufacture, and 

deployment of the manufactured 

reactor. 

The activities involving SNM within 

the manufacturing facility, including the 

loading of fuel, would be regulated 

primarily under the part 70 license. The 

reference to the requirements in subpart 

H of part 70 in section 53.620(d) assures 

that the activities involving the receipt, 

storage, and loading of a variety of 

possible fuel forms and enrichments at 

the manufacturing facility will be 

analyzed in a systematic manner and 

appropriate protection will be provided 

against equipment malfunctions, human 

errors, external hazards, and other 

adverse conditions. The regulations in 

part 51 provide a flexible approach for 

environmental review to address the 

range of regulated activities under part 

70. The flexibility in part 51 will enable 

the NRC to determine the appropriate 

type of environmental review based on 

the circumstances associated with the 

loading of fuel into a specific 

manufactured reactor. 

The proposed § 53.620(d) cites the 

requirements in parts 70, 71, and 73 to 

ensure important features and programs 

are in place prior to the receipt of SNM. 

The features and programs required to 

be in place prior to receipt of SNM 

include (1) radiation monitoring 

instrumentation and alarms; (2) 

measures to detect potential criticality 

accidents; (3) appropriate procedures, 

equipment, and personnel qualified for 

the fuel loading; (4) programs for 

physical security and cybersecurity; and 

(5) material control and accounting 

(MC&A) programs. Section 

53.620(d)(2)(i) proposes requirements to 

address security programs for any ML 

authorizing possession of a 

manufactured reactor into which fuel 

has been loaded at the manufacturing 

facility. Currently, for category II SNM, 

security measures may be required in 

addition to requirements included in 

§ 73.67, ‘‘Licensee fixed site and intransit requirements for the physical 

protection of special nuclear material of 

moderate and low strategic 

significance,’’ on a case-by-case basis. 

Including appropriate security measures 

in the proposed part 53 regulations will 

provide additional openness and 

transparency for applicants applying for 

an ML who seek to load fuel into 

manufactured reactors at a 

manufacturing site. 

Currently, § 73.67 only requires a 

security plan for licensees who possess, 

use, transport, or deliver to a carrier for 

transport SNM of moderate strategic 

significance, or 10 kg or more of SNM 

of low strategic significance. However, 

the proposed physical security program 

for fueled manufactured reactors would 

require a security plan for any ML 

authorizing possession of a 

manufactured reactor into which fuel 

has been loaded at the manufacturing 

facility, regardless of fuel type, 

enrichment, and quantity. This is 

consistent with other controls for MLs, 

including reactivity and criticality 

controls. 

The proposed requirements would 

also require a holder of an ML and part 

70 license to address cybersecurity to 

ensure a cyberattack would not 

adversely impact the functions 

performed by digital assets used by the 

licensee for physical security, radiation 

monitoring, or criticality prevention. 

The proposed regulations in part 53 

covering the activities related to the 

storage, movement, and loading of fresh 


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fuel into a manufactured reactor in the 

manufacturing facility would likewise 

refer to the applicable regulations in 

part 70. The proposed § 53.620(d) would 

also require the loading or unloading of 

unirradiated fuel into or from a 

manufactured reactor and any changes 

to the configuration of reactivity-related 

systems to be performed by a certified 

fuel handler meeting the requirements 

in subpart F. The NRC is aware of 

proposals to introduce reprocessing of 

existing or future spent nuclear fuel into 

the fuel cycle for some potential 

commercial nuclear plants. This 

proposed rule does not address the 

loading of spent nuclear fuel or fuel 

resulting from reprocessing of spent 

nuclear fuel into a manufactured 

reactor. 

Section 53.620(e) would limit the 

transport and delivery of a 

manufactured reactor or portions of a 

manufactured reactor only to a site for 

which the Commission has issued a 

COL authorizing the construction of a 

commercial nuclear plant using a 

manufactured reactor under the specific 

ML. This proposed requirement is 

similar to the limitations in § 52.153, 

with the difference being that part 53 

would allow the installation of a 

manufactured reactor at the site of a 

COL but would not include provisions 

for installation at a site under a CP. The 

possible combination of a manufactured 

reactor and the licensing option of CP 

and OL seems unlikely and would 

require the introduction of ITAAC into 

the licensing provisions for a CP and 

OL. An additional proposed paragraph 

in § 53.620(e) would provide 

requirements for protecting fueled 

manufactured reactors during transport 

to the site of the commercial nuclear 

plant by referencing the transportation 

and security requirements in 10 CFR 

part 71, ‘‘Packaging and Transportation 

of Radioactive Material,’’ and part 73. 

Section 53.620(f) would include 

proposed requirements for the 

acceptance and installation of a 

manufactured reactor at the site of a 

commercial nuclear plant. The proposed 

requirements would reference the 

construction requirements in § 53.610 to 

govern the integration of the 

manufactured reactor into the 

construction of a commercial nuclear 

plant. Other proposed requirements in 

the section would address required 

receipt inspections and verification that 

interface requirements between the 

manufactured reactor and the balance of 

the commercial nuclear plant have been 

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 § 53.710 would provide the 

requirements for maintaining 

capabilities, availability, and reliability 

of SSCs to demonstrate compliance with 

the safety criteria and design 

requirements for unplanned events that 

are described in proposed subparts B 

and C. The basic structure of this 

proposed section would be that controls 

for SR SSCs are provided by TS and 

controls for NSRSS SSCs are required to 

be addressed with licensee-controlled 

documents and procedures. 

The general content and control of TS 

under the proposed part 53 would be 

similar to the requirements in part 50. 

The proposed requirements for TS 

would include limits on the inventories 

of radioactive materials, plant operating 

limits, and specific requirements for 

each SR SSC, including limiting 

conditions for operation (LCO) and 

required surveillances. The proposed 

requirements for TS would also include 

a section on important design elements, 

which is similar to design features in 

§ 50.36, and a section for administrative 

controls. A provision addressing the 

development and submittal of TS to 

address decommissioning activities 

would also be included in the proposed 

subpart G. 

The proposed requirements for TS 

under part 53 would not carry over 

safety limits or associated limiting 

safety system settings from § 50.36, 

which contains TS requirements for 

operating reactors under parts 50 and 

52. As discussed in SECY–18–0096, 

systematic assessments and more 

mechanistic approaches to evaluating 

source terms support an alternative 

approach to establishing barrier-based 

safety limits. An example provided in 

that paper is a comparison of: (1) the 

traditional specified acceptable fuel 

design limits (SAFDL) that support 

protecting a specific barrier from 

potential failure mechanisms (e.g., 

departure from nucleate boiling to 

protect fuel cladding); and (2) the 

specified acceptable system 

radionuclide release design limit 

(SARRDL) concept, which limits the 

possible increase in circulating 

radionuclide inventory during normal 

operations or an AOO as part of an 

integrated or ‘‘functional containment’’ 

approach. Additional discussion of the 

use of SARRDL in the design and 

licensing of advanced reactors is 

provided in RG 1.232. The SARRDL 

could be addressed as an operating limit 

within this proposed construct of 

requirements for TS. In cases, such as 

LWRs, where a SAFDL approach might 

be used as part of a mechanistic 

approach to meeting the design and 

analysis requirements in subpart C, the 

associated functional design criteria 

proposed in § 53.410 and TS under the 

proposed § 53.710(a) would define 

similar requirements as those provided 

by the safety limit and limiting safety 

system setting requirements in § 50.36. 

The proposed requirements for TS 

under part 53 would not include 

specific criteria for identifying when 

LCOs must be established (i.e., would 

not include an equivalent to 

§ 50.36(c)(2)(ii)). Instead, consistent 

with subparts B and C, the TS 

requirements in subpart F of part 53 

would define TS LCOs as providing 

limits on SR SSCs. The SR SSCs protect 

against DBAs to demonstrate 

compliance with the safety criteria in 

the proposed § 53.210. In the proposed 

construct for part 53, risk-significant 

SSCs would be addressed through a 

combination of TS for the SR SSCs and 

establishment and monitoring of 

performance standards for NSRSS SSCs. 

In addition to addressing TS for SR 

SSCs, proposed § 53.710 would require 

appropriate controls be developed and 

implemented for NSRSS SSCs. 

Examples include appropriate 

surveillances and controls established 

through reliability assurance programs. 

Configuration management and other 

special treatments would provide that 

the capabilities, availabilities, and 

reliabilities of NSRSS SSCs are 

maintained consistent with the 

underlying risk assessments while 

providing flexibility to licensees 

through maintaining the management 

functions within licensee-controlled 

programs. Controls on NSRSS SSCs are 

appropriate as part of the overall 

performance-based approach within 

proposed part 53. Special treatments 

beyond those defined for their SR 

functions may also be warranted for SR 

SSCs to reflect their role in meeting the 

safety criteria in § 53.220 and the 

evaluation criteria in § 53.450(e). The 

performance objectives for NSRSS SSCs 

would reflect that the comprehensive 

risk metrics and related risk 

performance objectives established 

under § 53.220 may involve assessing 


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and averaging the risks over a defined 

period (e.g., plant year) and would not 

constitute a real-time requirement that 

must be continuously demonstrated by 

the licensee. The controls under 

§ 53.710(b) justify proposed changes in 

part 53 from the traditional or 

deterministic approaches in parts 50 

and 52 in areas such as replacing the 

single-failure criterion with a 

probabilistic reliability criterion (see 

SRM–SECY–03–0047, ‘‘Policy Issues 

Related to Licensing Non-Light-Water 

Reactor Designs,’’ dated June 26, 2003). 

This approach could also support the 

incorporation of risk insights and 

analytical margins to gain operational 

flexibilities in areas such as siting and 

staffing requirements described in 

subsequent sections of proposed subpart 

F. 

Proposed § 53.715 would provide the 

requirements for developing and 

implementing a program to do the 

following: (1) control maintenance 

activities; (2) take appropriate corrective 

action when performance issues are 

identified; (3) conduct routine 

evaluations of effectiveness; and (4) 

assess and manage risks resulting from 

maintenance activities. These proposed 

requirements are similar to those 

included in § 50.65 (maintenance rule), 

including the need to assess and manage 

the increase in risk that may result from 

the proposed maintenance activities. 

While, for the maintenance rule, 

specific criteria must be developed to 

capture both SR and non-SR but 

otherwise important SSCs, the proposed 

§ 53.715 would cover SR SSCs and 

NSRSS consistent with other subparts in 

part 53. 

Proposed § 53.720 would provide the 

requirements for responding to a 

seismic event during the operating 

phase of the life cycle of a commercial 

nuclear plant and would be equivalent 

to the requirements in paragraph 

IV(a)(3) of appendix S, ‘‘Earthquake 

Engineering Criteria for Nuclear Power 

Plants,’’ to part 50. 

The proposed part 53 would include 

provisions to address staffing, training, 

personnel qualifications, and human 

factors engineering (HFE) in a manner 

that is risk informed, technology 

inclusive, performance based, and 

flexible in nature. During the 

development of part 53, the staff 

prepared a draft white paper on ‘‘Risk 

Informed and Performance Based 

Human-System Considerations for 

Advanced Reactors,’’ to support 

interactions with stakeholders and the 

ACRS. Key considerations include the 

recognition that staffing, operator 

qualifications, and HFE are 

interconnected areas that must be 

approached in an integrated manner 

and, furthermore, that safety functions, 

including the means by which they are 

fulfilled, provide an effective method 

for informing technology-inclusive 

requirements. 

The requirements associated with this 

approach would be in §§ 53.725 through 

53.830. Section 53.725 discusses 

applicability and defines specific terms. 

Some definitions draw from those in 

§ 55.4. Several new definitions would be 

introduced for use within the context of 

subpart F. These new definitions would 

be the following: ‘‘Automation,’’ 

‘‘Auxiliary operator,’’ ‘‘Generally 

licensed reactor operator,’’ ‘‘Interactiondependent-mitigation facility,’’ ‘‘Load 

following,’’ ‘‘Self-reliant-mitigation 

facility.’’ 

Sections 53.725 through 53.830 would 

be divided into four portions that would 

cover general operational requirements, 

operator and senior operator licensing 

requirements, generally licensed reactor 

operator (GLRO) requirements, and 

general training requirements for plant 

staff. The NRC intends to provide 

guidance addressing the review of 

operator staffing plans; the review of 

operator, senior operator, and GLRO 

examination programs; and the 

implementation of scalable HFE 

reviews. Licensees would be required to 

use GLROs upon demonstrating 

compliance with the criteria in § 53.800. 

Certain routine communications are 

necessary to facilitate the operator 

licensing process. The NRC is proposing 

to adapt the requirements of §§ 55.5 and 

50.74 to § 53.726 to accomplish this. 

Specific information must be 

collected in order to facilitate the initial 

issuance of operator licenses, as well as 

to allow for license renewals and 

required updates thereafter. Such 

information collection activities must 

also be approved by the OMB. The NRC 

is proposing to adapt the requirements 

of § 55.8, to include any needed updates 

in OMB approval information, to 

§ 53.120 to accomplish this. 

The information used within the 

regulatory processes of the NRC must be 

free from omissions and inaccuracies to 

facilitate effective regulation. Consistent 

with this, the NRC is proposing to adapt 

the requirements of § 55.9 to § 53.728 to 

require the completeness and accuracy 

of material information provided by 

individual applicants and license 

holders. 

Section 53.730 would provide 

performance-based and technologyinclusive requirements for assessing the 

role of personnel in facility safety, 

applying human-system considerations 

within facility design, and incorporating 

operational approaches that are 

consistent with design-specific safety 

considerations. Most of these 

requirements would be adapted from 

portions of §§ 50.34(f) and 50.54 and 10 

CFR part 55, ‘‘Operators’ Licenses,’’ 

with considerable modification in order 

to reflect the introduction of new 

technologies and possible changes in 

the roles of personnel in preventing and 

mitigating events. The NRC is proposing 

that these technical requirements 

would, together, serve as a component 

of the required content of applications 

for OLs and COLs under part 53. 

Additionally, the NRC proposes that the 

specific technical requirements 

associated with HFE, human-system 

interface design, concept of operations, 

functional requirements analysis, and 

function allocation would serve as a 

component of the required content of 

applications for standard DCs, standard 

design approvals, MLs, and CPs, as well. 

Human factors engineering is 

essential to facilitate the role of 

personnel in facility safety in a manner 

that is both effective and reliable. The 

NRC proposes to adapt § 53.730(a) from 

the HFE design requirements of 

§ 50.34(f)(2)(iii). A key difference would 

be that the requirement would now be 

focused on settings where personnel 

fulfill their safety or emergency 

response roles wherever they may 

occur. The NRC additionally proposes 

to include within the scope of this 

requirement activities for assuring the 

continued availability of plant 

equipment that is needed for safety, and 

envisions that this may encompass 

relevant maintenance, inspections, and 

testing as well. The NRC intends that 

this requirement would be associated 

with staff guidance for conducting 

scalable reviews of HFE that is planned 

to accompany part 53. 

Human-system interfaces provide 

vital information to operators across a 

spectrum of operating conditions that 

can range from normal operations 

through severe accident conditions. The 

specific types of information that must 

be available to support operations staff 

during such conditions include, in part, 

those associated with safety function 

parameters, safety system status, 

possible core damage states, barrier 

integrity, and radioactive leakage. Due 

to the importance of such information, 

the NRC proposes under § 53.730(b) to 

require such human-system interface 

design features for all facilities, 

irrespective of other flexibilities 

proposed under part 53. Therefore, the 

NRC proposes to adapt specific postThree Mile Island requirements of 

§ 50.34(f) in a technology-inclusive 

manner as detailed in the following: 


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• Paragraph (b)(1) would be adapted 

from § 50.34(f)(2)(iv). 

• Paragraph (b)(2) would be adapted 

from § 50.34(f)(2)(v). 

• Paragraph (b)(3) would be adapted 

from § 50.34(f)(2)(xi), 50.34(f)(2)(xii), 

and 50.34(f)(2)(xxi). 

• Paragraph (b)(4) would be adapted 

from § 50.34(f)(2)(xvii), 50.34(f)(2)(xviii), 

50.34(f)(2)(xix), and 50.34(f)(2)(xxiv). 

• Paragraph (b)(5) would be adapted 

from § 50.34(f)(2)(xxvi). 

• Paragraph (b)(6) would be adapted 

from § 50.34(f)(2)(xxvii). 

In addition to the requirements of 

§ 53.730(b)(1) through (6), a further set 

of human-system interface design 

requirements applicable only to those 

facilities that will be staffed by GLROs 

would be provided under § 53.730(b)(7). 

This prescriptive set of design 

requirements for those facilities which 

demonstrate compliance with the 

criteria of § 53.800 would recognize that 

the application of HFE under § 53.730(a) 

is anticipated to be significantly 

reduced at such facilities in the absence 

of an expected operator role for the 

fulfillment of safety functions. However, 

it should be noted that the capability for 

an immediately initiated, manual 

reactor shutdown would be 

conservatively mandated irrespective of 

any other design considerations. 

The NRC proposes § 53.730(c) to 

require the submittal of a concept of 

operations that is of sufficient scope and 

detail to appropriately inform the staff. 

The development of a concept of 

operations can facilitate a clear 

understanding on the part of the NRC 

for potential novel operating concepts. 

Additionally, such information is likely 

to reduce the degree of resources and 

interactions needed for the NRC to 

obtain the understanding necessary to 

enable flexible requirements in areas 

such as staffing, operator qualifications, 

and HFE. 

The NRC proposes § 53.730(d) to 

require the submittal of both a 

Functional Requirements Analysis and a 

Function Allocation. The identification 

of design-specific safety functions and 

how they are fulfilled serves as a 

primary means for achieving 

technology-inclusive requirements 

within areas such as staffing, operator 

qualifications, and HFE. The Functional 

Requirements Analysis and Function 

Allocation processes (which are both 

HFE methods derived from systems 

engineering principles), provide an 

effective means to identify both how 

safety functions will be satisfied and 

how to characterize any associated 

operator role in doing so. A Functional 

Requirements Analysis shows what 

features, systems, and human actions 

are relied upon to demonstrate safety 

(i.e., fulfill safety functions). A Function 

Allocation then describes how safety 

functions are assigned to both personnel 

and automatic systems. However, an 

important adaptation of the Function 

Allocation for use under the proposed 

rule would be the further need to not 

only describe allocations of safety 

functions to human action and 

automation, but also to identify 

allocations made to active safety 

features, passive safety features, or 

inherent safety characteristics as well. 

Operating experience provides an 

important source of information by 

which to inform various aspects of 

facility design and operations. 

Accordingly, the NRC proposes in 

§ 53.730(e) to adapt the requirements of 

§ 50.34(f)(3)(i) for requiring an operating 

experience program. 

New technologies may involve 

concepts of operations that are more 

conducive to customizable licensed 

operator staffing requirements than the 

prescriptive requirements of § 50.54(m). 

Analyses and assessments that are based 

on HFE principles provide a 

performance-based means of 

determining licensed operator and 

senior operator staffing needed to 

support safe operations. In contrast, for 

those facilities required to be staffed by 

GLROs, the NRC anticipates that the 

operator staffing plans will reflect a 

simpler approach of showing that a 

continuity of responsibility will be 

maintained for facility operations 

throughout the operating phase, with at 

least one GLRO providing continuous 

oversight and remaining immediately 

available when any units are fueled. 

Additionally, a revised approach to the 

traditional position of the shift technical 

advisor that focuses on the availability 

of engineering expertise as a means of 

addressing uncertainties and abnormal 

circumstances is more suitable within 

the context of part 53 and is intended 

to be applicable to all facilities, 

irrespective of other design and staffing 

considerations. 

Consistent with this approach, the 

NRC proposes under § 53.730(f) to 

require the submittal of a staffing plan 

that details operations staffing, how 

engineering expertise will be provided, 

and what staffing will be available to 

provide other needed support functions. 

The NRC intends that this requirement 

would be associated with staff guidance 

for reviewing operations staffing plans 

that is planned to accompany part 53 

and that, following NRC approval of the 

OL or COL, the staffing plan would 

become a condition of the facility 

license. The NRC intends that, at a 

minimum, the approved licensed 

operator and senior operator (or, if 

applicable, GLRO) staffing, positions, 

and personnel locations will be 

incorporated into corresponding 

requirements within the facility TS and 

that a license amendment would thus be 

required for any subsequent changes. 

Operator training and qualification 

programs provide an essential 

component of supporting human 

performance in implementing tasks with 

safety implications. Such programs 

must include components that cover the 

stages of initial training, examination, 

and continuing training. Additionally, 

recognizing the potential for varying 

concepts of operations to affect 

traditional, prescriptive approaches to 

operator proficiency, the NRC proposes 

under part 53 to allow facilities to 

develop operator proficiency programs 

based on facility-specific 

considerations. 

Therefore, the NRC proposes in 

§ 53.730(g)(1) to require approval as part 

of its approval of the OL or COL, of the 

programs that will be used for the initial 

training, initial examination, 

requalification training and 

examination, and proficiency of both 

licensed operators and senior operators. 

In a corresponding manner, the NRC 

proposes in § 53.730(g)(2) to require 

approval of the programs that will be 

used for the GLRO equivalents of each 

of these programs for facilities with 

such staffing. The NRC intends that 

examination program requirements 

would be associated with staff guidance 

for the review of tailored examination 

processes that are planned to 

accompany part 53. Following the 

completion of an initial training 

program, continuing training programs 

provide an important means of 

sustaining the knowledge and abilities 

of individuals. The NRC is proposing to 

adapt the requirements of § 50.54(i–1) in 

§ 53.730(g)(3) to require that operator 

continuing training programs be in 

effect to support operator performance. 

Under part 53, the NRC proposes to 

require these programs to be in effect 

concurrent with when the initial 

operator examinations first commence, 

in effect putting the programs in place 

only when they are needed. This 

represents a modification of the 

comparable requirement of § 50.54(i–1), 

which links the commencement of these 

programs to a timeline driven by the 

licensing of the facility. 

The authorization to manipulate 

controls of the facility that directly 

affect reactivity or power level is 

restricted to individuals who are either 

licensed operators, licensed senior 

operators, or GLROs. However, for 

practical purposes, situations in which 


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an individual is participating in an 

approved training program or 

reestablishing proficiency may also call 

for them to operate the controls of the 

facility under the cognizance of a 

licensed individual. The NRC is 

proposing to adapt the requirements of 

§ 55.13 in § 53.735 to accomplish this, 

with a notable difference being the 

incorporation of GLROs. 

Section 53.740 would provide 

requirements for OL and COL holders 

under part 53. Portions of § 53.740 

would be adapted from the conditions 

of § 50.54. In general, the conditions for 

operations staffing under part 53 would 

reflect considerations for potential 

technological differences and varying 

concepts of operation that are expected 

among part 53 facility licensees. 

Additionally, certain requirements 

would be specific to the operating phase 

while others would remain in effect 

following the permanent cessation of 

facility operations during the 

decommissioning phase. 

All commercial nuclear plants 

licensed under part 53 would require 

some form of licensed operator staffing, 

whether it be by specifically or 

generally licensed operators. Consistent 

with this, the NRC is proposing under 

§ 53.740(a) to require facility licensees 

to demonstrate compliance with the 

programmatic requirements for either 

specifically licensed operators and 

senior operators or for GLROs, as 

applicable to the facility. 

The NRC recognizes that technologyinclusive facility staffing will need to 

account for a potentially wide range of 

concepts of operations; for this reason, 

flexible and performance-based 

approaches for establishing required 

facility staffing are appropriate. 

However, once the appropriate facility 

staffing has been determined and 

approved by the NRC, such staffing 

must be maintained to ensure that the 

appropriately qualified individuals will 

be available when needed to support the 

safe operation of the facility. Therefore, 

the NRC is proposing under § 53.740(b) 

to require that the staffing described 

within the approved facility staffing 

plan be maintained as a condition of the 

facility license as opposed to 

prescriptive staffing requirements like 

those of § 50.54(k) and (m). 

Because operation of facility controls 

directly affects reactivity or power level, 

only those individuals who possess 

appropriate levels of qualification and 

authorization are permitted to operate 

those controls. The NRC is proposing to 

adapt the requirements of § 50.54(i) in 

§ 53.740(c) to require that only 

specifically licensed operators and 

senior operators or, alternatively, 

GLROs, may operate facility controls, 

with allowance for specified exceptions 

for the purposes of operator training or 

proficiency. 

Senior operators, by virtue of their 

license level, are qualified and 

authorized both to perform certain 

important responsibilities and to direct 

the licensed activities of licensed 

operators. Therefore, facilities that are 

required to be staffed by specifically 

licensed operators must also include 

senior operators within their staffing. In 

contrast, facilities staffed with GLROs 

only have a single license level available 

and, therefore, there is no equivalent 

provision for such facilities. The NRC is 

proposing to adapt the requirements of 

§ 50.54(l) in § 53.740(d) to require the 

licensing and designation of senior 

operators at facilities staffed by 

specifically licensed operators. 

In contrast with control 

manipulations that directly affect 

reactor power and reactivity (e.g., 

control rod movement, control drum 

rotation, recirculation pump speed 

adjustment, reactor coolant system 

boration or dilution, etc.) and are 

therefore restricted to performance only 

by licensed operators, other types of 

plant operations that may result in 

reactor power and reactivity changes via 

means that are indirect in nature (e.g., 

electrical generation changes, turbine 

bypass valve operation, steam usage by 

process heat applications, etc.) may be 

implemented by non-licensed 

personnel. However, due to the 

potential influence of such operations 

on reactor power and reactivity, the 

continuous oversight of reactor 

parameters by a licensed operator is 

necessary during these operations. The 

NRC is therefore proposing to adapt the 

requirements of § 50.54(j) in § 53.740(e) 

to require appropriate oversight of 

operations, other than those associated 

with the controls themselves, that may 

affect reactivity or power level. 

Load following where plant output 

automatically changes in response to 

externally originated instructions or 

signals is not permitted under the 

existing regulations of § 50.54. However, 

new technological considerations and 

concepts of operation may justify such 

an operational approach under 

appropriate circumstances. The NRC 

recognizes that, beyond electrical power 

generation, load following may also 

affect other applications of plant output, 

such as hydrogen production, 

desalination, or district heating. For 

load following to be permissible, 

measures must be in place to provide 

assurance that plant output 

considerations are not permitted to lead 

to challenges to safe reactor operations. 

These measures may consist of 

automated control systems, automatic 

protective features, or the continuous 

oversight and immediate intervention 

capability of an appropriately qualified 

and authorized individual. Section 

53.740(f) would allow for load 

following, provided that appropriate 

measures are in place. In considering 

the acceptability of the measures 

associated with load following, the NRC 

expects that any automatic protection 

relied upon would be separate from that 

credited for reactor protection purposes 

and would employ setpoints that are set 

so as to prevent actuation of the reactor 

protection system while accomplishing 

its functions to the extent practical. 

Core alterations such as refueling are 

associated with specific considerations 

that warrant limiting the oversight of 

such operations to appropriately 

qualified and authorized individuals. 

Unlike other types of fuel handling 

operations, core alterations occur within 

the confines of a reactor vessel that is 

specifically designed to support and 

sustain nuclear criticality, thereby 

justifying the imposition of higher 

qualification levels within such 

contexts. The NRC is proposing to adapt 

the requirements of § 50.54(m)(2)(iv) in 

§ 53.740(g) to require the supervision of 

core alterations by either a specifically 

licensed senior operator, a specifically 

licensed senior operator whose license 

is limited to fuel handling, or by a 

GLRO, as applicable to the facility. 

Because certain commercial reactor 

designs may be capable of refueling 

while at power and, in any event, 

overall facility oversight would already 

be required by either a specifically 

licensed senior operator or by a GLRO, 

the NRC proposes to omit this 

requirement as redundant during 

periods where core alterations occur 

while the plant is operating. 

It is impossible to predict every 

possible scenario that a commercial 

nuclear plant might potentially 

encounter. Therefore, it is prudent to 

grant the authority for appropriately 

qualified individuals to depart from 

facility license conditions when 

emergency circumstances dictate that 

doing so is in the interest of public 

health and safety. The NRC is proposing 

to adapt the requirements of § 50.54(x) 

and (y) in § 53.740(h) to permit specific 

individuals to authorize departures from 

facility license conditions or TSs when 

emergency conditions warrant doing so 

for the protection of the public health 

and safety. Recognizing that certain 

facilities licensed under part 53 may be 

staffed by GLROs in lieu of specifically 

licensed senior operators, the NRC 

proposes to extend this authority to 


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GLROs. While it is not anticipated that 

GLROs will have a role in the 

fulfillment of safety functions at selfreliant-mitigation facilities and, 

furthermore, that operators at such 

facilities would not be in a position by 

which to significantly influence 

radiological safety outcomes, the very 

nature of the § 50.54(x) and (y) and the 

proposed § 53.740(h) provisions concern 

situations that are unanticipated and, 

therefore, unforeseeable. Thus, it is 

appropriate to grant GLROs a 

comparable authority to that of senior 

licensed operators and certified fuel 

handlers as it relates to invoking this 

provision under emergency conditions 

as a means of accounting for such 

possibilities. 

Due to the unique authorities and 

responsibilities of both specifically and 

generally licensed reactor operators, it is 

essential that any individual fulfilling 

such a role demonstrate compliance 

with the regulatory requirements for 

operator licensing. Section 107 of the 

Act authorizes the Commission to 

prescribe conditions for the licensing of 

operators and to issue licenses 

consistent with those conditions. The 

NRC is proposing to adapt the 

requirements of § 55.3 in § 53.745 to 

require that any person performing the 

function of an operator, senior operator, 

or GLRO must be authorized by a 

license issued by the Commission. 

The NRC proposes to license 

individuals as operators under both 

specific and general licensing 

frameworks. Specific licenses would be 

for licensed operators (i.e., reactor 

operators) and senior operators (i.e., 

senior reactor operators) and would be 

issued to a named person upon approval 

by the Commission of an application for 

that named person. In contrast, GLROs 

would perform duties under the 

provisions of a general license that 

would be effective without the filing of 

an application with the Commission or 

the issuance of licensing documents to 

a particular person. The NRC proposes 

requirements for the use of a specific 

licensing process for licensed operators 

and senior operators under §§ 53.760 

through 53.795, with § 53.760 

addressing applicability. 

Medical fitness is an important 

component of the overall process of 

specifically licensing operators because 

it provides assurance that operators will 

be able to carry out important duties 

without being precluded from doing so 

by health-related issues. Medical fitness 

also provides assurance that such issues 

will not adversely affect the 

performance of assigned job duties or 

cause operational errors that endanger 

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 

§§ 55.21, 55.23, and 55.27 under 

§ 53.765 to require medical fitness, 

examinations by physicians, and 

medical certification for specifically 

licensed operators and senior operators. 

In recognition of the fact that GLROs are 

not expected to have a role in the 

fulfillment of safety functions at the 

facilities at which they are licensed, the 

NRC proposes to not extend a 

comparable medical requirement to 

GLROs. 

The NRC is also proposing to adapt 

the requirements of §§ 55.25 and 

50.74(c) in § 53.770 to require that 

timely notifications be made to the NRC 

if a specifically licensed operator or 

senior operator develops a permanent 

physical or mental condition that 

adversely affects the performance of 

assigned operator job duties or could 

cause operational errors endangering 

public health and safety. 

Notwithstanding this requirement 

related to permanent medical 

conditions, the NRC continues to 

recognize that it is appropriate for 

facility licenses to impose 

administrative restrictions and 

conditions upon specifically licensed 

operators and senior operators in 

response to temporary medical 

conditions. 

The process of specifically licensing 

individuals as licensed operators or 

senior operators requires the submittal 

of applications to the NRC for review. 

These applications must detail certain 

elements associated with licensing, 

including the demonstration of 

compliance with examination, 

experience, and medical requirements. 

The NRC is proposing to adapt the 

requirements of §§ 55.31 through 55.35 

in § 53.775 to include requirements for 

the applications associated with the 

specific licensing of licensed operators 

and senior operators at commercial 

nuclear plants licensed under part 53. In 

contrast with the part 55 requirements, 

the NRC proposes to provide additional 

flexibility by locating certain details 

associated with the preparation and 

submittal of these applications within 

guidance in lieu of placement within 

this proposed rule itself. 

The NRC proposes overall 

programmatic requirements for 

specifically licensed operator and senior 

operator training, examination, and 

proficiency in § 53.780. In general, the 

proposed requirements are adapted from 

those in part 55, with several additional 

flexibilities being incorporated to better 

account for potential variations in 

reactor technologies and concepts of 

operations. The requirements proposed 

in § 53.780 cover, in part, the initial 

training, initial examination, 

requalification training, requalification 

examination, and proficiency of 

specifically licensed operators and 

senior operators. 

The initial training process provides 

individuals with the knowledge and 

abilities needed to subsequently fulfill 

assigned duties as licensed operators or 

senior operators in a safe and reliable 

manner. The use of a systems approach 

to training (SAT) ensures that the 

training program is based upon job 

requirements in a manner that can be 

adapted to account for differences in 

plant technology, concepts of 

operations, and operator roles in the 

fulfillment of design-specific safety 

functions. The NRC is proposing under 

§ 53.780(a) to require facility licensees 

to implement a SAT-based training 

program for the initial training of 

licensed operator and senior operator 

applicants. The program must be 

adequate to ensure that applicants will 

be capable of performing the duties 

necessary both to protect public health 

and safety and to maintain plant safety 

functions. The NRC further proposes 

that such programs be subject to NRC 

approval and subsequent change control 

processes of an appropriate nature. 

Examinations provide a means of 

assessing that individuals have achieved 

a degree of knowledge and ability that 

is sufficient to carry out assigned duties 

as licensed operators or senior operators 

in a manner that is safe and reliable. 

The NRC is proposing to adapt the 

requirements of §§ 55.40, 55.41, 55.43, 

and 55.45 in § 53.780(b) to require that 

facilities establish and implement an 

initial examination program. However, a 

key difference from the comparable 

requirements of part 55 would be that 

facilities have the flexibility to propose, 

subject to NRC approval, the 

examination methods and criteria to be 

used in assessing satisfactory applicant 

performance. Such examination 

programs (including those used within 

the scope of requalification training) 

would need to provide for acceptable 

levels of both test validity and test 

reliability in order to be considered 

acceptable. The NRC intends that staff 

guidance would be available to facilitate 

the review of licensing examination 

programs that are proposed by facility 

licensees and that, following NRC 

approval, initial examination programs 

would be subject to an appropriate 

change control process. Furthermore, 

the NRC proposes that holders of 

licenses to operate commercial nuclear 


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plants under part 53 be provided the 

alternative of administering their own 

approved licensing examinations. The 

NRC would continue to exercise 

appropriate oversight of the program, 

make operator licensing decisions based 

upon the examination results, and 

reserve the right to administer the 

examinations in lieu of permitting the 

facility to do so. However, irrespective 

of the provided flexibilities in 

examination format and structure, at a 

minimum, topics from the following 

general categories of knowledge and 

abilities should be sampled in such 

examinations: 

• Reactor Theory, Thermodynamics, 

and Chemical Interactions 

• Plant Systems and Components 

• Reactivity Management and 

Manipulations 

• Radiation Control and Safety 

• Emergency, Abnormal, and Normal 

Operations 

• Administrative Requirements and 

Conditions of the Facility License 

Requalification training programs 

provide for the continuing training and 

examination of specifically licensed 

operators and senior operators to ensure 

that they maintain the knowledge and 

abilities needed to support the safe and 

reliable performance of job duties 

following the completion of an initial 

training and examination program. The 

NRC is proposing to adapt the 

requirements of § 55.59 in § 53.780(c) to 

require that facilities implement both a 

SAT-based requalification training 

program and a biennial requalification 

examination program. However, a 

notable difference from the biennial 

requalification examinations required 

under part 55 would be that distinct 

annual operating test and biennial 

written examination components would 

not be mandated, with the facility 

licensee instead proposing the 

examination methods and criteria to be 

used in assessing satisfactory 

performance. The NRC intends that 

guidance would be available to facilitate 

the review of the requalification 

examination programs that are proposed 

by facility licensees and that, following 

NRC approval, requalification 

examination programs would be subject 

to an appropriate change control 

process. 

For examinations to provide for valid 

assessments of the knowledge and 

abilities of individuals, the 

examinations must remain free from 

compromises that could affect their 

underlying integrity. The NRC is 

proposing to adapt the requirements of 

§ 55.49 in § 53.780(d) to require that 

examinations and related activities 

remain free from any compromise that 

might affect the integrity of the 

examination process. 

Simulators provide a valuable means 

of training and evaluating plant 

operators, and the NRC is specifically 

authorized under the Nuclear Waste 

Policy Act of 1982, as amended 

(NWPA), section 306 (42 U.S.C. 10226) 

to establish regulations for the use of 

simulators within such context. The 

NRC is proposing to adapt the 

requirements of § 55.46 in § 53.780(e) to 

address the use of simulation facilities 

for training, examinations, and 

applicant experience requirements, as 

well as to address the maintenance of 

simulator fidelity. However, the 

proposed requirements of part 53 would 

not mandate that full scope, plantreferenced simulators be used and 

would allow the use of alternative 

simulation facilities consisting of, for 

example, partial scope simulators or the 

plant itself, provided that all associated 

requirements can be demonstrated to be 

met using alternative approaches and 

methods. Additionally, in allowing for 

the possibility that an applicant or 

licensee might demonstrate compliance 

with training, examination, or 

experience requirements using the plant 

itself, the NRC is not allowing the 

initiation of transients on the actual 

plant. Consistent with this, aside from 

controlled reactivity manipulations that 

are conducted for the purposes of 

demonstrating compliance with 

experience requirements, actual plant 

components may not be operated for 

these purposes. Rather, the NRC 

perspective is that the use of the plant 

for training and examination purposes 

should be restricted to techniques such 

as walkthroughs, job performance 

measures, simulated tasks, use of 

augmented reality technology, and 

similar approaches that provide training 

and examination value while avoiding 

the operation of actual plant 

components. 

There may be situations in which 

applicants for operator or senior 

operator licenses have previous training 

and experience that justifies waiving 

some, or all, of the initial examination 

requirements. The NRC is proposing to 

adapt the requirements of § 55.47 in 

§ 53.780(f) to allow for consideration of 

requests for waivers of examinations 

requirements. In contrast with the part 

55 requirements, the NRC proposes to 

locate certain details associated with 

such waiver requests within guidance 

documentation in lieu of placement 

within the rule itself. 

For licensed operators and senior 

operators to perform their assigned 

duties safely and reliably, it is essential 

that they perform those duties 

frequently enough so as to maintain a 

sufficient degree of proficiency. The 

NRC is proposing to adapt the 

requirements of § 55.53(e) and (f) in 

§ 53.780(g) to require that specifically 

licensed operators and senior operators 

maintain proficiency and, if proficiency 

is not maintained, regain proficiency 

prior to resuming licensed duties. 

However, in recognition of the fact that 

varying concepts of operations are 

possible for advanced reactor facilities, 

the NRC is proposing, in contrast with 

the requirements of part 55, to allow 

facility licensees to establish their own 

programs for operator proficiency, 

subject to NRC approval. 

As the holders of specific licenses, 

licensed operators and senior operators 

must be subject to license conditions on 

an individual basis to ensure that the 

basis upon which the licenses were 

issued remains valid. The NRC is 

proposing to adapt the requirements of 

§ 55.53 in § 53.785 to require 

appropriate conditions of licenses for 

specifically licensed operators and 

senior operators. However, in contrast 

with the requirements of § 55.53(e) and 

(f), the NRC is proposing to allow 

certain aspects of operator proficiency 

to be addressed by an NRC-approved 

facility proficiency program. 

Licenses for specifically licensed 

operators and senior operators are 

issued by the NRC and must remain 

subject to modification or revocation. 

The NRC is proposing to adapt the 

requirements of §§ 55.51 and 55.61 in 

§ 53.790 to address the issuance, 

modification, and revocation of licenses 

issued to specifically licensed operators 

and senior operators. 

The licenses issued to specifically 

licensed operators and senior operators 

are valid for a period of six years, after 

which they expire, unless otherwise 

renewed. The NRC is proposing to adapt 

the requirements of §§ 55.55 and 55.57 

in § 53.795 to address the expiration and 

renewal of licenses issued to 

specifically licensed operators and 

senior operators. 

In developing this proposed rule, the 

NRC has discussed with stakeholders 

the considerations that might justify the 

omission of the specifically licensed 

operators and senior operators. 

However, even for an inherently safe 

reactor with autonomous operation 

features, certain important 

administrative functions (e.g., 

compliance with TS, operability 

determinations, NRC notifications, 

emergency declarations, risk 

assessment, maintenance oversight, and 

radiological release limit compliance) 

would still need to be accomplished by 


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appropriately qualified and authorized 

individuals. Additionally, the NRC 

recognized that manual manipulations 

of facility reactivity controls must only 

be performed by individuals who have 

been appropriately licensed by the 

Commission. The NRC therefore 

proposes under § 53.800 to establish a 

new class of facility (defined as a selfreliant-mitigation facility), according to 

the criteria contained in § 53.800 for 

part 53. These facilities would employ 

GLROs rather than specifically licensed 

operators and senior operators. The 

GLRO regulations offer enhanced 

flexibilities and targeted relaxations in a 

manner that is commensurate with the 

modified role of such operators to 

ensure the safe operation of the 

associated facilities. In contrast, those 

facilities not meeting the criteria of 

§ 53.800 would instead be considered 

interaction-dependent-mitigation 

facilities and would require staffing by 

specifically licensed operators and 

senior operators. The terminology used 

to designate these facility types reflects 

differences in how operators are 

anticipated to need to interact with their 

plant systems in mitigating events and 

achieving safe outcomes; such systems 

may either need operators to interact 

with them in some manner (i.e., be 

interaction-dependent) or may instead 

be able to rely fully upon their own 

capabilities independent of operator 

interaction (i.e., be self-reliant). 

Generally licensed reactor operators 

would differ from specifically licensed 

operators because the latter would be 

directly and independently evaluated by 

the NRC as part of their licensing 

process. This direct and independent 

evaluation remains appropriate when 

operators may reasonably be expected to 

exert a significant influence on public 

health and safety outcomes. Therefore, a 

key determinant as to whether generally 

licensed reactor operators can be 

utilized in facility staffing is the 

assessment of the operator’s role in 

maintaining and fulfilling safety 

functions at the facility, such as through 

the performance of credited actions for 

the mitigation of plant events. 

The criteria proposed in § 53.800 

would designate self-reliant-mitigation 

facilities. These criteria are derived from 

the following set of considerations: 

• no human action needed to satisfy 

radiological consequence criteria; 

• no human action needed to address 

LBEs; 

• safety functions not allocated to 

human action; 

• reliance upon robust and highly 

reliable safety features; and 

• adequate defense in depth achieved 

without reliance on human action. 

It should be noted that those facilities 

not meeting the criteria proposed in 

§ 53.800 would instead be classified as 

interaction-dependent-mitigation 

facilities and would require staffing by 

specifically licensed operators and 

senior operators instead. 

Generally licensed reactor operators 

would perform duties under the 

provisions of a general license that 

would be effective without the filing of 

an application with the Commission or 

the issuance of licensing documents to 

a particular person. The NRC proposes 

requirements for the general licensing 

process for GLROs under §§ 53.805 

through 53.820. The requirements for 

GLROs would parallel those for senior 

operators in regard to their comparable 

administrative responsibilities. 

Nonetheless, the requirements for 

GLROs would be relaxed and 

incorporate greater flexibilities 

compared to the requirements for 

specifically licensed operators in a 

manner that is consistent with the 

GLRO’s role in safety at self-reliantmitigation facilities. 

In order to use GLROs in lieu of 

specifically licensed operators and 

senior operators, a OL/COL applicant 

would need to demonstrate that its 

proposed facility is a self-reliantmitigation facility, i.e., that it will 

comply with the following requirements 

on an ongoing basis: maintaining GLRO 

qualifications for the performance of 

important functions and tasks; 

incorporating relevant programmatic 

controls into TS; administering the 

related programs for training, 

examination, and proficiency; and 

ensuring that the relevant provisions of 

parts 26 and 73 are met. Additionally, 

to provide for an accurate accounting of 

what individuals are licensed under the 

general license, facility licensees would 

be required to report the identities of all 

generally licensed reactor operators to 

the NRC on an annual basis. 

Furthermore, a facility licensee must 

ensure that the facility design and 

performance continue to meet the 

technological criteria to be classified as 

a self-reliant-mitigation facility (i.e., the 

criteria of § 53.800) on a continual basis 

during the operating phase, as the 

relaxations afforded to such facilities in 

the areas of operator licensing, staffing, 

and HFE would be predicated on this 

assumption. The NRC therefore 

proposes under § 53.805 to establish 

requirements for facility licensees that 

address issues such as these. Finally, 

the failure of a self-reliant-mitigation 

facility to subsequently meet the criteria 

of § 53.800 after the issuance of an OL 

or COL would constitute a reportable 

event (i.e., an unanalyzed condition that 

significantly degrades plant safety) 

under the provisions of § 53.1630. 

The NRC proposes the general license 

for GLROs under § 53.810. GLROs 

would be licensed as a class of 

individuals under the provision of 

§ 53.810(a) and would be subject to the 

conditions specified in § 53.810(b) 

through (g). Portions of these conditions 

are adapted from § 55.53 and from those 

conditions currently included in the 

licenses issued to specifically licensed 

operators and senior operators. The NRC 

would retain the ability to suspend or 

prohibit individuals from operating 

under the general license should such 

action be warranted. 

The NRC proposes overall 

programmatic requirements for GLRO 

training, examination, and proficiency 

under § 53.815. In general, these 

proposed requirements are adapted from 

those of part 55 and parallel those also 

proposed for specifically licensed senior 

operators in § 53.780. These 

requirements include increased 

flexibilities and several targeted 

relaxations that reflect the limited role 

of GLROs in facility safety. The 

requirements proposed under § 53.815 

cover, in part, the initial training, initial 

examination, continuing training, 

requalification examination, and 

proficiency of GLROs. Section 53.805 

would require the facility licensee to 

develop, implement, and maintain these 

programs. Section 53.810, in turn, 

would prescribe that the requirements 

of § 53.805 would need to be met as a 

requirement of the general license. The 

implication of this structure is that the 

facility licensee would need to 

implement these programs for training, 

examination, and proficiency, and 

GLROs would need to participate in 

these programs to demonstrate 

compliance with the requirements of the 

general license. 

The initial training process provides 

GLROs with the knowledge and abilities 

needed to fulfill assigned duties as 

GLROs. The use of a SAT serves to 

ensure that the training program is 

based upon job requirements in a 

manner that can be adapted to account 

for differences in plant technology and 

concepts of operations. The NRC is 

proposing under § 53.815(b) to require 

facility licensees to implement a SATbased training program for the initial 

training of GLROs that is adequate to 

ensure that they have the necessary 

knowledge, skills, and abilities to 

perform their duties. The NRC further 

proposes that such programs would be 

subject to NRC approval, oversight, and 

appropriate change control processes. 

The training program must ensure that 


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GLROs maintain the necessary 

knowledge, skills, and abilities. 

Examinations provide a means of 

assessing that individuals have achieved 

a degree of knowledge and ability that 

will be sufficient to enable them to carry 

out assigned duties as GLROs in a 

manner that is both safe and reliable. 

The NRC proposes to adapt the 

requirements of §§ 55.40, 55.41, 55.43, 

and 55.45 in § 53.815(b) to require that 

facility licensees establish and 

implement an initial examination 

program. A key difference from the 

comparable requirements of part 55 

would be that facility licensees would 

be afforded the flexibility to propose, 

subject to NRC approval, the 

examination methods and criteria to be 

used in assessing satisfactory individual 

performance. Such examination 

programs (including those used within 

the scope of continuing training) would 

need to provide for acceptable levels of 

both test validity and test reliability in 

order to be considered acceptable. The 

NRC intends that staff guidance would 

be available to facilitate the review of 

initial examination programs that are 

proposed by facility licensees and that 

approved initial examination programs 

would be subject to an appropriate 

change control process. In contrast with 

both the requirements of part 55 and the 

proposed requirements of § 53.780, the 

NRC does not intend to administer or 

evaluate these initial examinations. 

However, the examination processes 

themselves will continue to be subject 

to ongoing NRC oversight. Irrespective 

of the provided flexibilities in 

examination format and structure, 

topics from the following general 

categories of knowledge and abilities 

should be sampled in such 

examinations: 

• Reactor Theory, Thermodynamics, 

and Chemical Interactions 

• Plant Systems and Components 

• Reactivity Management and 

Manipulations 

• Radiation Control and Safety 

• Emergency, Abnormal, and Normal 

Operations 

• Administrative Requirements and 

Conditions of the Facility License 

Continuing training programs provide 

the ongoing training and examination of 

GLROs to ensure that they maintain the 

knowledge and abilities needed to 

support the safe and reliable 

performance of job duties following the 

completion of an initial training and 

examination program. The NRC is 

proposing to adapt the requirements of 

§ 55.59 in § 53.815(b) to require that 

facility licensees implement both a 

SAT-based continuing training program 

and a requalification examination 

program. However, a notable difference 

from the examinations required under 

part 55 would be that distinct annual 

operating test and biennial written 

examination components would not be 

mandated. The facility licensee would 

instead propose examination methods 

and criteria to be used in assessing 

satisfactory performance. Furthermore, 

unlike the comparable requirements of 

part 55 and those proposed for 

specifically licensed operators and 

senior operators, a biennial periodicity 

for requalification examinations would 

not be prescribed. However, adequate 

justification for the proposed periodicity 

of requalification examinations would 

be required. The NRC intends that staff 

guidance would be available to facilitate 

the review of the requalification 

examination programs that are proposed 

by facility licensees. Approved 

requalification examination programs 

would be subject to an appropriate 

change control process. 

For examinations to provide for valid 

assessments of the knowledge and 

abilities of individuals, the 

examinations must remain free from 

compromises that could affect their 

underlying integrity. The NRC is 

proposing to adapt the requirements of 

§ 55.49 in § 53.815(d) to require that 

examinations and related activities 

remain free from any compromise that 

might affect the integrity of the 

examination process. 

Simulators provide a valuable means 

of training and evaluating plant 

operators and the NRC is specifically 

authorized under the NWPA, section 

306 (42 U.S.C. 10226) to establish 

regulations for the use of simulators 

within such context. The NRC is 

proposing to adapt the requirements of 

§ 55.46 in § 53.815(e) to address the use 

of simulation facilities for training and 

examinations, and experience 

requirements, as well as to address the 

maintenance of simulator fidelity. The 

use of full scope, plant-referenced 

simulators would not be mandated. The 

potential use of alternative simulation 

facilities consisting of, for example, 

partial scope simulators or the plant 

itself, would be allowed provided that 

all associated requirements could be 

demonstrated to be met using 

alternative approaches and methods. 

Additionally, in allowing for the 

possibility that an applicant or licensee 

might demonstrate compliance with 

training and examination requirements 

using the plant itself, the NRC is not 

allowing the initiation of transients on 

the actual plant. Consistent with this, 

aside from controlled reactivity 

manipulations that are conducted for 

the purposes of demonstrating 

compliance with experience 

requirements, actual plant components 

may not be operated for these purposes. 

Rather, the use of the plant for training 

and examination purposes should be 

restricted to techniques such as 

walkthroughs, job performance 

measures, simulated tasks, use of 

augmented reality technology, and 

similar approaches that provide training 

and examination value while avoiding 

the operation of actual plant 

components. 

There may be situations in which 

GLROs have previous training and 

experience that justifies waiving some, 

or all, of the initial examination. 

Therefore, the NRC is proposing under 

§ 53.815(f) to allow facility licensees to 

waive some, or all, portions of initial 

examinations provided that such 

waivers are consistent with a program 

that has been approved by the NRC. 

For GLROs to safely and reliably 

perform their assigned duties, it is 

essential that they perform those duties 

frequently enough so as to maintain a 

sufficient degree of proficiency. 

However, the NRC recognizes that 

facilities that utilize GLROs may have 

concepts of operation that warrant 

unique proficiency considerations. 

Therefore, the NRC is proposing in 

§ 53.815(g) to require that facility 

licensees develop, implement, and 

maintain programs to maintain and 

reestablish, if needed, the proficiency of 

GLROs. This could occur, for example, 

if an individual’s extended absence 

from watch standing has rendered 

proficiency requirements unmet. 

The general license should remain in 

effect for an individual only while that 

individual remains employed in a 

position that may call for the individual 

to manipulate the reactivity controls of 

the facility. The NRC proposes under 

§ 53.820 to require that the general 

license would cease to be applicable on 

an individual basis when an 

individual’s employment status 

becomes such that this is no longer the 

case. However, the NRC recognizes that 

for some types of self-reliant-mitigation 

facilities, very long periods may elapse 

between circumstances that necessitate 

manual manipulation of reactivity 

controls. Therefore, the general license 

remains in effect for an individual as 

long as the individual’s current position 

could potentially require that individual 

to manipulate reactivity controls at 

some point within the course of the 

individual’s assigned job duties. 

The NWPA, section 306 (42 U.S.C. 

10226) authorizes and directs the NRC 

to, in part, issue regulations and 

guidance that address the training and 


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qualifications of civilian nuclear power 

plant operators, supervisors, 

technicians, and other appropriate 

operating personnel. The NRC 

implements this in part 50 through the 

requirements of § 50.120, ‘‘Training and 

qualification of nuclear power plant 

personnel.’’ The NRC is proposing 

under § 53.830 to adapt, with 

modifications, the requirements of 

§ 50.120 for use in part 53 to provide 

more flexible personnel training and 

qualification requirements than those in 

§ 50.120 and better reflect diverse 

concepts of operations. 

The NRC recognizes that the 

categories of nuclear power plant 

personnel in § 50.120 may not be 

needed for the diverse concepts of 

operations, staffing models, and nontraditional personnel roles and 

responsibilities anticipated under 

proposed part 53; conversely, and for 

the same reasons, additional categories 

of plant personnel may need to be 

covered by part 53. The NRC also 

recognizes that the timeframe prescribed 

in § 50.120 for the establishment of 

training programs may not be aligned 

with the schedules associated with the 

startup of certain types of commercial 

nuclear plant facilities. However, the 

NRC also recognizes that the SAT-based 

training required under § 50.120 

remains an appropriate means by which 

training programs should continue to be 

developed and implemented. Therefore, 

the approach taken by the NRC in 

addressing the training of certain plant 

staff under the proposed part 53 reflects 

greater flexibilities in personnel 

categories and programmatic 

timeframes, while still retaining the 

requirement that such training programs 

be based on SAT. 

The NRC is proposing under § 53.830 

to require SAT-based training programs 

with the timeframe for when such 

programs are required being based upon 

when the associated personnel are 

needed to support facility-specific 

needs. The training programs would 

cover the training and qualification of 

plant personnel in the general categories 

of supervisors, technicians, and other 

appropriate operating personnel. The 

licensee would not be required to seek 

NRC approval of a training program 

prior to usage. However, the licensee is 

required to accommodate NRC 

inspection of the training program. The 

NRC intends to develop guidance to 

facilitate the inspection of these training 

programs but does not intend for such 

guidance to preclude the potential for 

the training programs to be maintained 

by a separate, NRC-approved 

accreditation process. 

The proposed § 53.845 would require 

programs to be developed, 

implemented, and maintained to help 

ensure that design features and human 

actions have the capabilities and 

reliabilities necessary to demonstrate 

compliance with the safety criteria in 

subpart B throughout the operating life 

of each commercial nuclear plant. The 

proposed programmatic requirements in 

subpart F would also address areas such 

as radiation protection needed to 

control routine effluents during normal 

operations. The proposed §§ 53.850 

through 53.910 would require programs 

to support specific activities needed to 

ensure the prevention or mitigation of 

unplanned events or to support normal 

operations for any reactor design. 

However, each holder of an OL or COL 

would be required to assess whether 

additional programs are needed for the 

specific reactor design and location of 

the commercial nuclear plant. Licensees 

would be able to combine, separate, and 

otherwise organize programs and related 

documents as appropriate for the 

technologies and organizations 

associated with the commercial nuclear 

plant. 

Proposed § 53.850 would require a 

radiation protection program associated 

with the requirements in subparts B and 

C for public doses resulting from normal 

operations and the protection of plant 

workers. The proposed requirements 

related to doses from normal operations, 

including routine effluents, would be 

similar to those specified in § 50.36a, 

‘‘Technical specifications on effluents 

from nuclear power reactors,’’ and 

related requirements in standard TS for 

offsite dose calculation manuals. While 

the proposed section would include 

requirements that are technically and 

programmatically similar to part 50, 

proposed § 53.850 would not include a 

requirement for effluent-related TS as is 

required in § 50.36a. A proposed 

requirement similar to that found in the 

administrative controls section of TS for 

operating reactors licensed under parts 

50 and 52 would be included for 

programmatic controls of solid wastes to 

complement the design requirements in 

proposed § 53.425. 

Proposed § 53.855 would require an 

emergency response plan that 

demonstrates compliance with the 

requirements in appendix E to part 50 

and § 50.47(b) or § 50.160. The 

regulations in § 50.47 stating that the 

NRC will not issue certain licenses 

unless it finds that there is reasonable 

assurance that adequate protective 

measures can and will be taken to 

protect public health and safety in the 

event of a radiological emergency apply 

equally to applications under part 53 

complying with the applicable 

standards set forth in either § 50.160 or 

the requirements in appendix E to part 

50 and § 50.47(b). 

In its 2008 Advanced Reactor Policy 

Statement, the Commission stated their 

expectation that ‘‘the safety features of 

advanced reactor designs will be 

complemented by the operational 

program for Emergency Planning (EP). 

This EP operational program, in turn, 

must be demonstrated by inspections, 

tests, analyses, and acceptance criteria 

to ensure effective implementation of 

established measures.’’ Consistent with 

this policy statement, emergency plans 

and emergency planning zones are not 

safety features in the design. In SECY– 

97–020, ‘‘Results of Evaluation of 

Emergency Planning for Evolutionary 

and Advanced Reactors,’’ dated January 

27, 1997, the staff indicated that the 

rationale upon which EP for current 

reactor designs is based, that is, 

potential consequences from a spectrum 

of accidents, is appropriate for use as 

the basis for EP for evolutionary and 

passive advanced LWR designs and is 

consistent with the Commission’s 

defense-in-depth safety philosophy. 

Also, in its Safety Goals Policy 

Statement the Commission stated that: 

‘‘A defense-in-depth approach has been 

mandated in order to prevent accidents 

from happening and to mitigate their 

consequences. Siting in less populated 

areas is emphasized. Furthermore, 

emergency response capabilities are 

mandated to provide additional defensein-depth protection to the surrounding 

population.’’ Consistent with this policy 

statement, proposed § 53.855 

contributes an additional independent 

layer of defense in depth for commercial 

nuclear plants. Therefore, the 

emergency plans and emergency 

planning zones under proposed § 53.855 

are not used to demonstrate compliance 

with subpart B and subpart C of this 

part. Rather, compliance with the 

requirements in proposed § 53.855 

would provide reasonable assurance 

that adequate protective measures can 

and will be taken to protect public 

health and safety in the event of a 

radiological emergency. 

Proposed § 53.860 would identify the 

applicable regulations for part 53 

applicants related to the programs for 

physical security, cybersecurity, FFD, 

AA, and information security. These 

programs are discussed in more detail in 

section V, ‘‘Changes to Other Parts of 10 

CFR,’’ of this document. 

Proposed § 53.860(a) would establish 

the physical protection program and 

present a graded approach to physical 

protection requirements. If a licensee 

can meet the proposed criterion in 


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2SECY–22–0072, ‘‘Proposed Rule: Alternative 

Physical Security Requirements for Advanced 

Reactors,’’ dated August 2, 2022. 

§ 53.860(a)(2)(i), then the requirement to 

protect against the design-basis threat 

(DBT) of radiological sabotage would 

not be applicable. The criterion in 

§ 53.860(a)(2)(i) would require a 

licensee to show that potential 

consequences resulting from a DBT 

initiated event would result in offsite 

doses below the values in § 53.210 even 

if licensee mitigation and recovery 

actions, including any operator action, 

are unavailable or ineffective. Where the 

criterion is met, the resulting physical 

protection requirements would be those 

for protection of SNM and Category 1 

and Category 2 radioactive material, if 

applicable. This proposal would apply a 

new regulatory approach for certain 

commercial nuclear plants in which the 

DBT of radiological sabotage would not 

be applicable. 

For those licensees able to meet the 

criterion in § 53.860(a)(2), the NRC 

would not conduct Force-On-Force 

(FOF) exercise inspections. Section 

170D.a of the Act permits the 

Commission to determine which 

licensed facilities are part of a class of 

licensed facilities where NRCconducted FOF exercises are 

appropriate to assess the ability of a 

private security force of a licensed 

facility to defend against any applicable 

DBT. For the class of licensees that meet 

the criterion of § 53.860(a)(2), it would 

not be appropriate to conduct FOF 

exercises to evaluate performance at 

commercial nuclear plants where the 

DBT of radiological sabotage is not 

applicable and the facility poses a lower 

risk to public health and safety from 

potential radiation exposure. These 

facilities would still have tailored 

security requirements and oversight 

consistent with their relatively low risk. 

For those licensees not able to meet 

the criterion in § 53.860(a)(2), proposed 

§ 53.860(a) would permit the licensee to 

choose one of two paths to provide 

physical protection: (1) the current set 

of requirements in § 73.55, which would 

include any changes resulting from the 

ongoing proposed rulemaking on 

Alternative Physical Security 

Requirements for Advanced Reactors 2

that provides pre-determined physical 

security alternatives; or (2) the 

performance-based requirements in 

proposed § 73.100. In either case, the 

licensee would be subject to NRCconducted FOF inspections. 

Proposed § 53.860(b) would require 

licensees to establish, implement, and 

maintain an FFD program under part 26. 

Section 53.860(c) would require 

licensees to establish, implement, and 

maintain an AA program in accordance 

with either § 73.56 or proposed § 73.120, 

as appropriate. Section 53.860(d) would 

require licensees to establish, 

implement, and maintain a 

cybersecurity program in accordance 

with either § 73.54 or proposed § 73.110. 

Section 53.860(e) would require 

licensees to establish, implement, and 

maintain an information protection 

system that complies with the 

requirements of §§ 73.21, 73.22, and 

73.23, as applicable. 

Proposed § 53.865 would establish 

requirements for quality assurance and 

refer to appendix B of part 50 for the 

part 53 requirements for SR design 

features. Proposed requirements related 

to evaluating and reporting changes to 

the quality assurance program would be 

included in proposed subpart I and 

would be equivalent to those found in 

§ 50.54. 

The proposed § 53.870 would require 

licensees to actively assess possible 

degradation of SSCs from the effects of 

aging, fatigue, and environmental 

conditions. The proposed inclusion of 

requirements related to designing and 

monitoring for possible degradation 

mechanisms reflects important lessons 

learned from the history of LWRs and 

the likely introduction of new design 

features and materials in future 

commercial nuclear plants. The 

allowable combinations of design 

features, operating experience, testing, 

and monitoring during operations 

would support performance-based 

approaches to the initial licensing of 

new technologies. The proposed 

performance-based approach to integrity 

assessment programs would also allow 

for the subsequent consideration of 

operating experience and appropriate 

corrective actions or allowable 

relaxations for ensuring that design 

features comply with the proposed 

functional design criteria of §§ 53.410 

and 53.420. The proposed program 

would be based upon a comprehensive 

and integrated evaluation of the aging 

and other degradation mechanisms 

applicable to the design; identification 

of the affected SSCs; the allowances 

provided in the design of the SSCs for 

degradation; and schedules and 

procedures for determining if and at 

what rate degradation is occurring, as 

well as its cause. Risk insights could be 

used to prioritize the monitoring, 

evaluation, and management of 

degradation based upon the importance 

of the SSC to safety and the time frame 

for when the effects of degradation 

could be of concern. 

Proposed § 53.875 would establish 

requirements for a fire protection 

program supporting operations similar 

to § 50.48. The proposed fire protection 

program during operations would work 

in concert with specific fire protection 

requirements proposed in subpart C for 

design and analyses and in proposed 

subpart E for construction and 

manufacturing. 

Proposed § 53.880 would establish 

requirements for an inservice inspection 

(ISI) and inservice testing (IST) program, 

which are historically important 

activities conducted in accordance with 

ASME codes and regulations in 

§ 50.55a. While the proposed part 53 

would not incorporate specific 

consensus codes and standards into the 

regulations, § 53.880 allows for the use 

of generally accepted codes and 

standards. The proposed requirement 

for an ISI and IST program would 

reinforce the need to develop 

monitoring programs to be conducted 

during a plant’s operations phase to 

complement the design process and 

address inherent uncertainties. The NRC 

encourages the continued use of 

consensus codes and standards 

supporting design, testing, and 

inspections to support integrated and 

performance-based approaches in 

demonstrating compliance with the 

proposed requirements in part 53. 

Proposed § 53.910 would establish 

requirements for developing, 

implementing, and maintaining 

procedures (e.g., operations and 

emergency operating procedures) and 

guidelines (e.g., accident management 

guidelines). The programmatic 

requirements for many of the 

procedures listed in this proposed 

section would be similar to the 

requirements found in the 

administrative controls section of TS for 

plants licensed under parts 50 and 52. 

The proposed inclusion, where 

appropriate, of accident management 

guidelines in these requirements is 

intended to ensure that an integrated set 

of procedures and guidelines would be 

established by licensees to ensure 

command and control across the 

spectrum of possible event sequences. 

The proposed required procedures 

would also include those needed to 

complement the design requirements in 

proposed § 53.440(m) related to 

criticality alarms and the equivalent of 

the procedures required in § 50.54(hh) 

to address notifications of potential 

aircraft threats. 

Subpart G—Decommissioning 

Requirements 

The proposed subpart G would 

provide the regulatory requirements for 

the decommissioning phase of the life 

cycle of a commercial nuclear plant. 


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The requirements being proposed in 

subpart G for the decommissioning of a 

commercial nuclear plant are adapted 

from the current regulations in § 50.75, 

‘‘Reporting and recordkeeping for 

decommissioning planning,’’ § 50.82, 

‘‘Termination of license,’’ and § 50.83, 

‘‘Release of part of a power reactor 

facility or site for unrestricted use.’’ 

Although the requirements from those 

sections of part 50 have been copied 

into proposed subpart G with relatively 

few changes, the requirements are 

reorganized to fit within the part 53 

structure. The few changes made were 

primarily to make the proposed 

requirements more technology inclusive 

by adding alternatives within sections, 

whereas some requirements in part 50 

were developed specifically for LWRs. 

As an example, § 50.75 provides 

minimum amounts of decommissioning 

funds required to demonstrate 

reasonable assurance of funds for 

decommissioning LWRs. Such generic 

amounts have not been developed for all 

reactor technologies that may be 

licensed under part 53. Therefore, the 

Commission proposes in § 53.1020, 

‘‘Cost estimates for decommissioning,’’ 

that site-specific cost estimates for 

decommissioning must be developed 

considering costs in such areas as 

engineering, labor, and waste disposal. 

The derivation of the generic cost 

estimates for LWRs in § 50.75 is 

provided in NUREG/CR–5884, ‘‘Revised 

Analyses of Decommissioning for the 

Reference Pressurized Water Reactor 

Power Station,’’ and NUREG/CR–6187, 

‘‘Revised Analyses of Decommissioning 

for the Reference Boiling Water Reactor 

Power Station.’’ Similar to part 50, a 

provision for an annual adjustment of 

decommissioning cost estimates would 

be included in proposed § 53.1030. 

The NRC is currently pursuing 

another rulemaking, ‘‘Regulatory 

Improvements for Production and 

Utilization Facilities Transitioning to 

Decommissioning,’’ which was 

published as a proposed rule for public 

comment on March 3, 2022 (87 FR 

12254). As these rulemakings progress, 

the NRC will consider revisions to part 

53 to align the two rulemaking efforts. 

For example, the proposed § 53.1075 

could be expanded to include or 

reference requirements for 

decommissioning in areas such as EP 

and security in addition to the proposed 

decommissioning fire protection plans 

that would provide an equivalent to 

§ 50.48(f). 

Subpart H—Licenses, Certifications, and 

Approvals 

Proposed subpart H would provide 

requirements related to applications 

under part 53 for NRC licenses, 

certifications, or approvals for 

commercial nuclear plants. 

Proposed subpart H would specify 

requirements applicable to all part 53 

applications as well as requirements 

specific to part 53 applications for 

LWAs, ESPs, standard design approvals, 

standard DCs, MLs, CPs, OLs, and COLs. 

Proposed subpart H would be 

equivalent to and include all existing 

licensing, certification, and approval 

processes currently covered under parts 

50 and 52, with the exception of the 

process for early review of site 

suitability issues. Interactions with 

external stakeholders during the 

development of the proposed rule did 

not identify significant interest in or 

need for including the process for early 

review of site suitability issues in part 

53. 

Much of the proposed subpart H 

regulatory text is identical to the 

corresponding language in parts 50 and 

52, with minor changes to account for 

cross references in part 53, to make 

language technology neutral, or to 

reflect the unique analytical approach in 

part 53. In these instances, this 

preamble discussion will describe the 

language as ‘‘equivalent’’ to the existing 

corresponding requirement in part 50 or 

part 52 and will describe any 

deviations, where applicable. 

Because part 53 carries over the 

majority of the licensing options from 

parts 50 and 52, there are several 

sections in proposed subpart H that are 

similar to existing regulations in parts 

50 and 52. Proposed § 53.1100 would 

address filing of applications for 

licenses, certifications, or approvals 

under oath or affirmation and is 

equivalent to § 50.30. The proposed 

§ 53.1100 does not include the current 

requirement in § 50.30(a)(2) that the 

applicant maintain the capability to 

generate additional copies, because it is 

unnecessary in the age of electronic 

submissions. In addition, the existing 

requirement on applications for OLs in 

§ 50.30(d) is included in proposed 

§ 53.1124(g)(2), ‘‘Relationship between 

sections,’’ covering OLs, rather than in 

proposed § 53.1100. 

Proposed § 53.1101 would lay out 

activities requiring an NRC license and 

is equivalent to § 50.10(b). Proposed 

§ 53.1103 would address combining 

applications and is equivalent to 

§§ 50.31, 50.52, and 52.8. Proposed 

§ 53.1103(b) would continue the 

Commission’s practice of combining 

multiple authorizations for a facility 

under parts 30, 40, 50, 52, and 70 into 

one license based on the Commission’s 

authority under Section 161h. of the Act 

to combine NRC licenses. Proposed 

§ 53.1106 would address elimination of 

repetition and is equivalent to § 50.32. 

Proposed § 53.1109 would provide 

general information requirements for the 

content of applications submitted to the 

NRC under part 53 and is equivalent to 

§ 50.33, with the exception of § 50.33(f) 

on financial qualifications, which is 

covered in proposed subpart J, and 

§ 50.33(h) on earliest and latest dates for 

completion of construction, which is 

covered in § 53.1306 of this subpart. 

Each application would need to include 

information to address the items in 

proposed § 53.1109 as cited in the 

appropriate section of this subpart for 

the application type. 

One change from current 

requirements can be found in proposed 

§ 53.1109(i), which is not limited to 

electricity generation as it is currently in 

part 50. Some prospective NRC 

applicants are considering development 

of nuclear plants for other commercial 

ventures, such as process heat 

generation or hydrogen production. In 

addition, § 53.1109(j), which requires 

applications containing classified 

information to separate that information 

from the unclassified information in the 

application, refers to ‘‘Restricted Data or 

classified National Security 

Information’’ instead of the term used in 

the corresponding provision in 

§ 50.33(j), ‘‘Restricted Data or other 

defense information.’’ This change was 

made to use the defined term in part 95 

rather than ‘‘defense information’’ as 

used in § 50.33(j). The usage in § 50.33(j) 

dates back to the Atomic Energy 

Commission amendment of that section 

on January 19, 1956 (21 FR 355, 357) 

and was not changed with the issuance 

of part 95 (45 FR 14476; March 5, 1980) 

after the establishment of the NRC and 

the 1975 reissuance of the former 

Atomic Energy Commission regulations. 

The revised terminology also aligns 

with its usage in § 53.1115. 

Proposed § 53.1112 would address 

environmental conditions and is 

equivalent to § 50.36b. Proposed 

§ 53.1115 would address requirements 

for agreements limiting access to 

classified information and is equivalent 

to § 50.37. Proposed § 53.1118 would 

address ineligibility of certain 

applicants and is equivalent to § 50.38. 

Proposed § 53.1120 would address 

exceptions and exemptions from 

licensing requirements for Department 

of Defense and DOE facilities and is 

equivalent to § 50.11. Proposed 

§ 53.1121 would address public 

inspection of applications and is 

equivalent to § 50.39. 

Proposed § 53.1124 would address the 

relationship between the various 

licenses, certifications, and approvals 


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provided in this subpart, and the 

requirements are equivalent to a number 

of similar provisions in parts 50 and 52 

including §§ 50.10, 52.13, 52.43, 52.73, 

52.133, and 52.153. New provisions are 

provided in § 53.1124(c) and (d), that 

would allow an application for either a 

standard design approval or a standard 

DC under part 53 to reference applicable 

licensing-basis information that 

supported issuance of an OL or COL 

under part 53. These provisions are 

being proposed to offer additional 

flexibility beyond what is currently 

allowed under parts 50 or 52 for an 

applicant who may wish to license a 

first-of-a-kind reactor for operation prior 

to seeking generic approval or 

certification of the standard design. 

Proposed § 53.1124(e) would address 

the limitations that a manufactured 

reactor may only be transported to a site 

with a COL and is equivalent to 

§ 52.153. Proposed § 53.1130 would 

address LWAs and is equivalent to 

§ 50.10. 

Proposed §§ 53.1140 through 53.1188 

would govern the content of ESP 

applications. Proposed § 53.1140 is 

equivalent to § 52.12. Proposed 

§ 53.1143 would address filing of 

applications and is equivalent to 

§ 52.15. Proposed § 53.1144 would 

address general information 

requirements for the content of 

applications and is equivalent to 

§ 52.16. 

Proposed § 53.1146 would specify 

requirements for the technical contents 

of applications and is equivalent to 

§ 52.17. Proposed § 53.1146(b)(2) 

provides applicants for ESPs a 

regulatory option to propose major 

features of the emergency plans or 

complete and integrated emergency 

plans in accordance with either the 

requirements in § 50.160 of this chapter, 

or the requirements in appendix E to 

part 50 of this chapter and § 50.47(b) of 

this chapter, as applicable. 

Proposed § 53.1149 would address 

standards for review of ESP applications 

and administrative review of 

applications, including hearings, and is 

equivalent to §§ 52.18 and 52.21. 

Proposed § 53.1155 would address 

referral to the ACRS and is equivalent 

to § 52.23. Proposed § 53.1158 would 

address issuance of ESPs and is 

equivalent to § 52.24. Proposed 

§ 53.1161 would address the extent of 

activities permitted and is equivalent to 

§ 52.25. Proposed § 53.1164 would 

address the duration of an ESP and is 

equivalent to § 52.26. Proposed 

§ 53.1167 would address provisions for 

requesting a LWA after issuance of an 

ESP and is equivalent to § 52.27. 

Proposed § 53.1170 would address 

transfers of ESPs and is equivalent to 

§ 52.28. Proposed § 53.1173 would 

address applications for ESP renewals 

and is equivalent to § 52.29. Proposed 

§ 53.1176 would address criteria for 

renewal of an ESP and is equivalent to 

§ 52.31. Proposed § 53.1179 would 

address the duration of an ESP renewal 

and is equivalent to § 52.33. Proposed 

§ 53.1182 would address the use of a 

site for purposes other than those 

described in the permit and is 

equivalent to § 52.35. Proposed 

§ 53.1188 would address finality of ESP 

determinations and is equivalent to 

§ 52.39. 

Proposed §§ 53.1200 through 53.1221 

would govern the contents of standard 

design approval applications. Proposed 

§ 53.1200 is equivalent to § 52.131. 

Proposed § 53.1203 would address filing 

of applications and is equivalent to 

§ 52.135. Proposed § 53.1206 would 

address general information 

requirements for the content of 

applications and is equivalent to 

§ 52.136. 

Proposed § 53.1209 would address 

requirements for the technical content 

of applications and is largely equivalent 

to § 52.137. In proposed § 53.1209(a), 

the NRC proposes text that expands the 

discussion of ‘‘major portion’’ standard 

design approvals. Additional discussion 

regarding standard design approvals for 

a major portion of a standard design can 

be found in the NRC’s ‘‘A Regulatory 

Review Roadmap for Non-Light Water 

Reactors,’’ which considers the Nuclear 

Innovation Alliance report ‘‘Clarifying 

‘Major Portions’ of a Reactor Design in 

Support of a Standard Design 

Approval.’’ Proposed § 53.1209(b) 

outlines the required content of the 

Final Safety Analysis Report (FSAR). 

Proposed requirements in 

§ 53.1209(b)(2) for portions of the 

application addressing design 

information state that the application 

must include design information 

equivalent to that required for a 

standard DC. This reference to the 

pertinent DC requirements (specifically, 

those in proposed § 53.1239(a)(2) 

through (27)) is an efficiency that would 

prevent the need to repeat many of the 

same requirements for the content of a 

standard design approval application. 

Proposed § 53.1210 would address 

requirements for the content of a 

standard design approval application 

other than the FSAR. Proposed 

§ 53.1210(a) would require the inclusion 

of a description of availability controls 

that are not included in the FSAR. 

Proposed § 53.1212 would address 

standards for review of applications and 

is equivalent to § 52.139. Proposed 

§ 53.1215 would address referral to the 

ACRS and is equivalent to § 52.141. 

Proposed § 53.1218 would address staff 

approval of designs and duration of 

design approvals and is equivalent to 

§§ 52.143 and 52.147. Proposed 

§ 53.1221 would address finality of 

standard design approvals and 

information requests and is equivalent 

to § 52.145 with the exception that it 

extends such finality to a standard 

approval referenced in a DC application. 

Standard design approvals issued to 

date under part 52 have been issued 

during the NRC’s review of the standard 

DC application and have relied on the 

same application content. However, a 

future scenario could arise where the 

DC application is not submitted until 

after a design approval has been 

granted. The NRC would apply the same 

finality provisions in this situation as in 

the situation where a standard design 

approval is referenced in a COL 

application. 

There is no equivalent to proposed 

§ 53.1221(d) in part 52 for standard 

design approvals. This provision would 

state that the Commission will require, 

before granting a CP, COL, OL, or ML 

which references a standard design 

approval, that engineering documents 

be completed and available for audit. A 

similar provision is included in part 52 

in relation to a standard DC; and the 

NRC would require that design and 

analysis information needed for the 

Commission to make its safety 

determination be complete and 

available for any application the NRC is 

reviewing. Making this explicit provides 

increased clarity to future standard 

design approval applicants under part 

53. 

Proposed §§ 53.1230 through 53.1263 

would address standard DCs. Proposed 

§ 53.1230 would address general 

provisions for standard DCs and is 

equivalent to § 52.41. Proposed 

§ 53.1233 would address filing of 

applications and is equivalent to 

§ 52.45. Proposed § 53.1236 would 

address general information 

requirements for the content of 

applications and is equivalent to 

§ 52.46. Proposed § 53.1239 would 

address requirements for the technical 

content of applications and is 

equivalent to § 52.47(a). The 

requirements in proposed § 53.1239 

have been modified from the analogous 

requirements in § 52.47(a) to align with 

the technical requirements in proposed 

part 53. 

Proposed § 53.1241 would address 

requirements for the content of a 

standard DC application other than the 

FSAR and is equivalent to § 52.47(b) 

and (d). 


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Proposed § 53.1242 would address 

review of applications and is equivalent 

to §§ 52.48 and 52.51. Proposed 

§ 53.1242(c) would include a provision 

that would allow a DC applicant to 

reference applicable licensing-basis 

information for an OL or COL issued 

under part 53. As explained previously, 

this provision is being proposed to 

explicitly allow flexibility for an 

applicant who may wish to license a 

first-of-a-kind reactor for operation prior 

to seeking certification of the generic 

reactor design. For NRC findings on a 

reactor design in an OL or COL 

proceeding, this proposal would 

provide finality in a subsequent DC 

application that references information 

on the OL or COL proceeding’s docket. 

This finality accorded to the OL or COL 

findings would bind the NRC staff and 

the ACRS but would not bind members 

of the public or the Commission. (To the 

extent an Atomic Safety and Licensing 

Board (ASLB) might have a role in a DC 

rulemaking, the OL or COL findings 

would not bind the ASLB either.) 

Specifically, members of the public 

would have the opportunity to comment 

on a proposed DC rule under wellestablished NRC practice. The rationale 

for binding the NRC staff and ACRS is 

similar to the rationale for a COL 

applicant referencing a standard design 

approval under part 52. 

Proposed § 53.1245 would address 

referral to the ACRS and is equivalent 

to § 52.53. Proposed § 53.1248 would 

address issuance of standard DCs and is 

equivalent to § 52.54. Proposed 

§ 53.1251 would address duration of 

certifications and is equivalent to 

§ 52.55(c). Proposed § 53.1254 would 

address application for renewal and is 

equivalent to § 52.57. Proposed 

§ 53.1257 would address criteria for 

renewal and is equivalent to § 52.59. 

Proposed § 53.1260 would address 

duration of renewals and is equivalent 

to § 52.61. Proposed § 53.1263 would 

address finality of standard DCs and is 

equivalent to § 52.63. 

Proposed §§ 53.1270 through 53.1291 

would address MLs covering 

manufacturing activities at one or more 

licensee facilities. Proposed § 53.1270 

would address the scope of these 

sections and is equivalent to § 52.151. 

Proposed § 53.1273 would address 

filing of applications for an ML and is 

equivalent to § 52.155(a). 

Proposed § 53.1276 would address 

general information requirements for the 

content of ML applications and is 

equivalent to § 52.156, with one 

exception. Proposed § 53.1276 would 

require each application for an ML to 

also include the information required by 

§ 53.1109(e). This information includes 

the type of license applied for, the use 

to which the facility will be put, the 

period of time for which the license is 

sought, and a list of other licenses, 

except operator’s licenses, issued or 

applied for in connection with the 

proposed facility to address the 

potential variations in how MLs might 

be formulated under the proposed part 

53. 

Proposed § 53.1279 would address 

requirements for the technical content 

of applications for MLs to be included 

in the FSAR and is equivalent to 

§ 52.157. In addition, the requirements 

in proposed § 53.1279(a) and (b) have 

been modified from the analogous 

requirements in § 52.157 to align with 

the technical requirements in proposed 

part 53. Proposed § 53.1279(a)(2) 

outlines the required content of the 

application addressing design 

information and states that the 

application must include design 

information equivalent to that required 

for a standard DC. This reference to the 

pertinent DC requirements is an 

efficiency that would prevent the need 

to repeat the same requirements for the 

content of an ML application. 

Proposed § 53.1279(c) would provide 

application requirements related to the 

deployment of the completed 

manufactured reactor. Proposed 

§ 53.1279(c)(1) would require inclusion 

of information related to the procedures 

governing the preparation of the 

manufactured reactor for shipping to the 

site where it is to be operated, the 

conduct of shipping, and the 

verification of the condition of the 

shipped items upon receipt at the site. 

Proposed § 53.1279(c)(2) would require 

that the application include information 

on the interaction of the design, 

manufacture, and installation of a 

manufactured reactor within the 

applicant’s organization and the manner 

by which the applicant will ensure close 

integration between the designer, 

contractors, and any licensee of a 

facility in which the manufactured 

reactor is to be installed. Finally, 

proposed § 53.1279(c)(3) would require 

that the application include a 

description of the measures used for the 

control of interfaces between the holder 

of the ML and the holder of the COL for 

the commercial nuclear plant at which 

the manufactured reactor is to be 

installed. This information is necessary 

for the NRC to determine whether the 

applicant would have appropriate 

controls in place to ensure coordination 

between parties involved in the design, 

manufacture, and eventual operation of 

any reactor manufactured under an ML. 

Proposed § 53.1279(d) would include 

additional requirements for application 

content for applicants seeking an ML for 

manufactured reactors that will be 

fueled at the factory under a 10 CFR part 

70 license, consistent with the 

requirements in § 53.620(d). These 

provisions would require the 

application to include information 

related to loading fuel and the required 

independent physical mechanisms to 

prevent criticality and to otherwise 

provide assurance that the fueled 

manufactured reactor can be 

successfully transported, installed, and 

operated at a site for which the 

Commission has issued a COL that 

authorizes construction and operation of 

a commercial nuclear plant using the 

manufactured reactor. 

Proposed § 53.1282 would provide 

requirements for other application 

content for MLs and is equivalent to 

§ 52.158. Proposed § 53.1282(a)(1) 

would provide requirements to include 

in the ML application the ITAAC within 

the scope of the ML that the COL holder 

referencing the ML must satisfy. 

Proposed § 53.1282(a)(2) would require 

that the ITAAC from a referenced 

standard design apply to the portions of 

the ML design within the scope of the 

referenced standard design. Proposed 

§ 53.1282(a)(3) would state that the COL 

application may include a notification 

that required referenced standard DC 

ITAAC have been satisfied at the 

manufacturing facility. 

Proposed § 53.1282(b) would require 

an ML application to include an 

environmental report and, consistent 

with existing requirements, proposed 

§ 53.1282(b)(2) would note that if the 

ML application references a standard 

DC, the environmental report need not 

contain a discussion of severe accident 

mitigation design alternatives for the 

manufactured reactor as used in a 

commercial nuclear plant. 

Proposed § 53.1285 would provide 

standards for review of applications and 

administrative review of applications 

for MLs, including hearings, and is 

equivalent to §§ 52.159 and 52.163. 

Proposed § 53.1286 would address 

referral of applications to the ACRS and 

is equivalent to § 52.165. Proposed 

§ 53.1287 would address issuance of an 

ML and is equivalent to § 52.167. 

Proposed § 53.1288 would address 

finality of MLs and is equivalent to 

§ 52.171. Proposed § 53.1291 would 

address the duration of MLs and is 

equivalent to § 52.173. Proposed 

§ 53.1293 would address the transfer of 

MLs and is equivalent to § 52.175. 

Proposed § 53.1295 would address the 

renewal of MLs and is equivalent to 

§§ 52.177, 52.179 and 52.181, with a 

minor exception. Proposed 

§ 53.1295(a)(3) would state that an ML 


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for which a timely application for 

renewal has been filed remains in effect 

until the Commission has made a final 

determination on the renewal 

application, provided, however, that the 

holder of an ML may not begin 

manufacture of a manufactured reactor 

less than six months before the 

expiration of the license. The proposed 

6-month time frame for this provision is 

changed from the 3-year period in the 

equivalent provision in part 52 because 

future reactor applicants may present 

smaller, simpler designs, to include 

micro-reactor designs, in ML 

applications than those that were 

envisioned when the existing 

requirements were written. A 6-month 

time frame for this provision would 

provide greater flexibility for ML 

holders related to manufactured reactors 

being produced when the ML expires. 

Proposed §§ 53.1300 through 53.1348 

would address licensing requirements 

for CPs. Proposed § 53.1300 would set 

out general requirements for CPs and is 

equivalent to § 50.23. Proposed 

§ 53.1306 would address the general 

information requirements for the 

content of applications for CPs and is 

equivalent to § 50.33(f) and (h). 

Proposed § 53.1309 would address 

requirements for the technical content 

of applications for CPs and includes the 

requirement to submit a Preliminary 

Safety Analysis Report (PSAR) that 

describes the facility and presents a 

preliminary safety analysis of the 

facility as a whole. This is in contrast to 

an OL application which is required to 

include an FSAR that describes the 

facility and presents a final safety 

analysis of the facility as a whole. 

Proposed § 53.1309 is equivalent to 

§ 52.17(a)(1)(iv) through (a)(1)(x) and 

52.17(b), with two exceptions. First, 

proposed § 53.1309 would replace the 

analysis of the dose criteria required by 

§ 52.17(a)(1)(ix) with analysis to 

demonstrate compliance with the safety 

criteria defined in §§ 53.210 and 53.220. 

Second, proposed § 53.1309(a)(2) would 

add a requirement for a CP application 

to include several categories of detailed 

design information, although 

§ 53.1309(a)(2)(ii) would allow certain 

relaxations of this requirement in view 

of aspects of a design that may not yet 

be fully developed. Section 53.1309 

would reference the requirements for 

the content of an ESP application to 

address application requirements 

related to siting and would reference the 

requirements for the content of a DC 

application to address application 

requirements related to design of the 

commercial nuclear plant. Proposed 

§ 53.1309(a)(2)(ii) would address the 

treatment of preliminary design 

information and notes that information 

provided in the application may include 

some aspects of the design that are not 

fully developed. This provision would 

require that the completed design, 

including any changes during 

construction, be described in the FSAR 

in an application for an OL. This would 

include the requirement for a 

description of the PRA required by 

§ 53.450(a) and its results. Probabilistic 

risk assessments developed for 

commercial nuclear plants prior to 

construction would be based on the 

design and other information available 

at the time of the CP application. PRAs 

performed in early design stages or prior 

to construction may be inherently less 

detailed and may include projected 

information that will be subsequently 

verified or revised when the plant is 

built. Proposed § 53.1309(a)(4) would 

address preliminary description of the 

plans for coping with emergencies. 

Proposed § 53.1312 would address 

other application content for CPs. 

Proposed § 53.1312(a)(1) is equivalent to 

§ 52.80(b) but is adapted for a CP 

application. Proposed § 53.1312(a)(2) is 

equivalent to § 52.80(c) but is adapted 

for a CP application. Proposed 

§ 53.1312(b)(1) is equivalent to 

§ 52.79(b), (c), and (d) but is adapted for 

a CP application. Section 53.1312(b)(2) 

is equivalent to portions of 

§§ 52.63(b)(1), 52.79(b)(1) through (b)(3), 

(c), and (d)(1) and (d)(3), 52.80, and 

52.93(b), but is adapted for a CP 

application. Guidance for equivalent 

requirements in parts 50 and 52 is also 

addressed in RG 1.206, ‘‘Applications 

for Nuclear Power Plants,’’ Revision 1, 

section C.1.7. 

Proposed § 53.1315 would address 

standards for review of applications and 

administrative review of applications, 

including hearings, and is equivalent to 

§§ 52.81 and 52.85, but is adapted for a 

CP application. 

Proposed § 53.1318 would address 

finality of NRC approvals, licenses, and 

certifications referenced in a CP 

application and is equivalent to 

§ 52.83(a) but is adapted for a CP 

application. 

Proposed § 53.1324 would address 

referral to the ACRS and is equivalent 

to § 50.58(a) and to § 52.87 but is 

adapted for a CP application. 

Proposed § 53.1327 would address 

authorization to conduct LWA activities 

and is equivalent to § 52.91 but is 

adapted for a CP application. Proposed 

§ 53.1327(a) is equivalent to § 52.91(a) 

but is adapted for a CP application. 

Proposed § 53.1327(b) is equivalent to 

§ 52.91(b) but is adapted for a CP 

application. Proposed § 53.1330 would 

address exemptions, departures, and 

variances for CP applicants. 

Proposed § 53.1333 would address 

issuance of CPs. Proposed § 53.1333(a) 

is equivalent to § 50.35(a). Proposed 

§ 53.1333(b) is equivalent to § 50.35(b) 

and to § 52.97(c) but is adapted for a CP 

application. Proposed § 53.1336 would 

address the effect of CPs and is 

equivalent to § 50.35(b). Proposed 

§ 53.1342 would address the duration of 

CPs. Proposed § 53.1342(a) is equivalent 

to § 50.55(a). Proposed § 53.1342(b) is 

equivalent to § 50.55(b). Proposed 

§ 53.1345 would address the transfer, 

assignment, and disposal of CPs and is 

equivalent to § 50.80. Proposed 

§ 53.1348 would address the 

termination of CPs and is equivalent to 

§§ 52.3(b)(8) and 52.110(a)(1) but is 

adapted for a CP application. 

Proposed §§ 53.1360 through 53.1405 

address requirements for OLs. 

Proposed § 53.1366 would address 

requirements for the general content of 

applications for OLs. It would refer to 

general content requirements in 

proposed § 53.1109 and would require 

supplemental information. Proposed 

§ 53.1366(a) is equivalent to § 50.33(f). 

Proposed § 53.1366(b) is equivalent to 

§ 50.33(k). 

Proposed § 53.1369 would provide 

requirements for the technical content 

of applications for OLs to be included 

in the FSAR and is equivalent to 

§ 50.34(b) but has been modified to align 

with the technical requirements in part 

53. It would require that the FSAR 

include and, as needed, update 

information provided in the PSAR that 

was submitted and reviewed to support 

the associated CP application. 

Similar to the proposed requirements 

for the content of CP applications, 

proposed § 53.1369(a) would reference 

the requirements for the content of an 

ESP application to address application 

requirements related to the site. Section 

53.1369(b) would reference the 

requirements for the content of a DC 

application to address some of the 

application requirements related to 

design of the commercial nuclear plant. 

Proposed § 53.1369(c) is equivalent to 

§ 50.34(b)(7). Proposed § 53.1369(d) 

would require a description of the 

Integrity Assessment Program that 

would be required by proposed 

§ 53.870. Proposed § 53.1369(e) is 

equivalent to § 50.34(e). Proposed 

§ 53.1369(g) would provide 

requirements for OL application content 

to support proposed § 53.730 related to 

the role of personnel in the operation of 

the commercial nuclear plant and is 

adapted from requirements in part 55 

and § 50.34(f). Likewise, proposed 

§ 53.1369(h) would provide 


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requirements for OL application content 

related to training programs to support 

proposed §§ 53.730(g) and 53.830 and 

includes requirements equivalent to 

§ 50.34(b)(8), § 52.79(a)(33), and part 55. 

Proposed § 53.1369(i) would provide 

requirements for OL application content 

related to emergency plans to support 

proposed § 53.855 and is equivalent to 

§ 50.34(b)(6)(v). 

Proposed § 53.1369(j) would provide 

requirements for OL application content 

related to the applicant’s organizational 

structure and is equivalent to 

§ 50.34(b)(6)(i). Proposed § 53.1369(k) 

would provide requirements for OL 

application content related to the 

applicant’s proposed maintenance 

program to support proposed § 53.715 

and is equivalent to § 50.34(b)(6)(iv). 

Proposed § 53.1369(l) would provide 

requirements for OL application content 

related to the applicant’s quality 

assurance program to support proposed 

§ 53.865 and is equivalent to 

§ 50.34(b)(6)(ii). Proposed § 53.1369(m) 

would provide requirements for OL 

application content related to the 

applicant’s proposed radiation 

protection program to support proposed 

§ 53.850 and is equivalent to 

§ 50.34(b)(3). 

Proposed § 53.1369(n) through (p) 

would provide requirements for OL 

application content related to the 

applicant’s proposed physical security 

program to support proposed § 53.860(a) 

and are equivalent to § 50.34(c) and (d). 

Proposed § 53.1369(q) would provide 

requirements for OL application content 

related to the applicant’s proposed 

cybersecurity plan to support proposed 

§ 53.860(d) and is equivalent to 

§§ 52.79(a)(36)(iv) and 73.54. Proposed 

§ 53.1369(r) would provide 

requirements for OL application content 

related to the implementation of 

proposed security, safeguards, and 

cybersecurity plans to support proposed 

§ 53.860 and is equivalent to 

§ 52.79(a)(35)(ii) and 52.79(a)(36)(iv) 

and (v). 

Proposed § 53.1369(s) would provide 

requirements for OL application content 

related to the applicant’s proposed fire 

protection program to support proposed 

§ 53.875 and is equivalent to 

§ 52.79(a)(40). Proposed § 53.1369(t) 

would provide requirements for OL 

application content related to the 

applicant’s proposed ISI and IST 

program to support proposed § 53.880 

and is equivalent to part of 

§ 52.79(a)(11). Proposed § 53.1369(w) 

would provide requirements for OL 

application content related to the 

applicant’s general employee training 

program to support proposed § 53.830 

and is equivalent to § 52.79(a)(33). 

Proposed § 53.1369(x) would provide 

requirements for OL application content 

related to the applicant’s FFD program 

to support part 26 and is equivalent to 

§ 52.79(a)(44). Proposed § 53.1369(y) 

would provide requirements for OL 

applicant’s programs to demonstrate 

that any safety questions identified at 

the CP stage have been resolved and is 

equivalent to § 50.34(b)(5). Proposed 

§ 53.1369(z) would provide 

requirements for OL applicants to 

describe how the performance of each 

safety design feature has been 

demonstrated capable of fulfilling 

functional design criteria considering 

interdependent effects through either 

analysis, appropriate test programs, 

prototype testing, operating experience, 

or a combination thereof to support 

proposed § 53.440(a). It is largely 

equivalent to §§ 50.34(b)(5) and 

50.43(e). Proposed § 53.1369(aa) would 

provide requirements for OL application 

content related to the applicant’s 

proposed TS to support proposed 

§ 53.710(a) and is equivalent to 

§ 50.34(b)(6)(vi). 

Proposed § 53.1372 would address 

requirements for the content of OL 

applications other than the FSAR. 

Proposed § 53.1372(a) would require 

submission of an environmental report 

and is equivalent to § 50.30(f) and 

§ 51.53(b). Proposed § 53.1372(b) does 

not have a direct parallel in parts 50 and 

52 and would require the inclusion of 

a description of availability controls that 

are not included in the FSAR to support 

proposed § 53.710(b). 

Proposed § 53.1375 would address 

standards for review of OL applications 

and the administrative review of 

applications, including hearings, and is 

equivalent to §§ 52.81 and 52.85, except 

that the NRC has omitted 10 CFR part 

54, ‘‘Requirements for Renewal of 

Operating Licenses for Nuclear Power 

Plants,’’ from the list of standards in the 

proposed § 53.1375(a). Proposed part 53 

does not include detailed requirements 

related to renewal of licenses, although 

a general provision and possible 

placeholder for future requirements has 

been included as proposed § 53.1595. 

The NRC will decide after the part 53 

final rule is published whether this 

future section will be retained in part 53 

to address license renewal or whether 

the agency will take another approach to 

address license renewal for part 53 

licensees, such as amending part 54 to 

address part 53 licensees. 

Proposed § 53.1381 would address 

referral to the ACRS and is equivalent 

to §§ 50.58 and 52.87. Proposed 

§ 53.1384 would address exemptions, 

departures, and variances for OL 

applicants. Section 53.1384(a) is 

equivalent to § 52.93 but is adapted for 

OLs. Proposed § 53.1384(b) is equivalent 

to §§ 52.39(d) (with respect to ESPs) and 

52.93 but is adapted for OLs. 

Proposed § 53.1387 would address 

issuance of OLs. The proposed 

introductory paragraph is equivalent to 

§ 50.56. Proposed § 53.1387(a)(1)(i) is 

equivalent to §§ 50.50 and 50.57(a)(1). 

Proposed § 53.1387(a)(1)(ii) is 

equivalent to § 50.50. Proposed 

§ 53.1387(a)(1)(iii) is equivalent to 

§ 50.57(a)(2). Section 53.1387(a)(1)(iv) is 

equivalent to § 50.57(a)(3). Proposed 

§ 53.1387(a)(1)(v) is equivalent to 

§ 50.57(a)(4). Proposed 

§ 53.1387(a)(1)(vi) is equivalent to 

§ 50.57(a)(6). Proposed 

§ 53.1387(a)(1)(vii) is equivalent to 

§ 50.57(a)(5). Proposed 

§ 53.1387(a)(1)(viii) is equivalent to 

§ 52.97(a)(1)(vi) but is adapted for OLs. 

Proposed § 53.1387(c) is equivalent to 

§ 50.57(b). Proposed § 53.1387(d) is 

equivalent to §§ 50.36(b) and 50.50. 

Proposed § 53.1390 would address 

backfitting of OLs and is equivalent to 

§ 52.98(a) but adapted for an OL 

application. Proposed § 53.1396 would 

address duration of an OL and is 

equivalent to § 50.51(a) and § 52.104. 

Proposed § 53.1399 would address 

transfer, assignment, and other 

disposition of an OL and is equivalent 

to § 50.80. Proposed § 53.1402 would 

address applications for renewal of an 

OL and refers to proposed § 53.1595. 

Proposed § 53.1405 would address 

continuation of an OL and is equivalent 

to § 52.109 but is adapted to address an 

OL. 

Proposed §§ 53.1410 through 53.1461 

would address requirements for COLs. 

Proposed § 53.1410 is equivalent to 

§ 52.71. Proposed § 53.1413 would 

address general information 

requirements for the content of 

applications for COLs and is equivalent 

to § 52.77, which references § 50.33. 

Most of the provisions from § 50.33 are 

restated in proposed § 53.1109. Some 

requirements in § 50.33 related to 

financial qualifications and construction 

timelines are addressed in other 

sections of part 53. 

Proposed § 53.1416 would address the 

technical content to be included in an 

FSAR for an application for a COL and 

is equivalent to § 52.79 except as 

modified to reflect the technical 

requirements in part 53 and with one 

addition. Proposed § 53.1416 includes 

the statement that the Commission will 

require, before issuance of a COL, that 

engineering documents, such as 

analyses, drawings, procurement 

specifications, or construction and 

installation specifications, be completed 

and available for audit if the more 


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detailed information is necessary for the 

Commission to verify the information in 

the application and make its safety 

determination. This statement is 

equivalent to DC application 

requirements in § 52.47 and is included 

in proposed § 53.1416 for clarity. 

Similar to the proposed requirements 

for the content of OL applications, 

proposed § 53.1416(a)(1) would 

reference the requirements for the 

content of an ESP application to address 

application requirements related to 

siting. Section 53.1416(a)(2) would 

reference the requirements for the 

content of a DC application to address 

some of the application requirements 

related to design of the commercial 

nuclear plant. The remaining items 

under proposed § 53.1416(a) are 

likewise similar to the required content 

for OL applications under proposed 

§ 53.1369(a). Proposed § 53.1416(b) 

would require COL applicants to 

provide a report documenting the 

resolution of any safety questions for 

SSCs for which research and 

development was necessary to confirm 

the adequacy of their design and is 

equivalent to § 50.34(b)(5). Proposed 

§ 53.1416(c) would provide 

requirements for COL applicants to 

describe how the performance of each 

safety design feature has been 

demonstrated capable of fulfilling 

functional design criteria considering 

interdependent effects through either 

analysis, appropriate test programs, 

prototype testing, operating experience, 

or a combination thereof to support 

proposed § 53.440(a). It is largely 

equivalent to §§ 52.79(a)(24) and 

50.43(e). Proposed § 53.1416(d) would 

address the content of COL applications 

referencing an ESP. Proposed 

§ 53.1416(e) would address the content 

of COL applications referencing a 

standard design approval. Proposed 

§ 53.1416(f) would address the content 

of COL applications referencing a 

standard DC. Proposed § 53.1416(g) 

would address the content of COL 

applications referencing an ML. 

Proposed § 53.1419 would address 

other application content for COLs and 

is equivalent to § 52.80. Proposed 

§ 53.1419(a)(2) is new and would 

require the inclusion of a description of 

availability controls that are not 

required to be included in the FSAR. 

Proposed § 53.1422 would address 

standards for review of applications and 

the administrative review of 

applications, including hearings, and is 

equivalent to §§ 52.81 and 52.85. The 

NRC has removed part 54 from the list 

of standards in proposed § 53.1422(a). 

Proposed part 53 does not include 

requirements related to renewal of 

licenses, in relation to proposed 

§§ 53.1422 and 53.1595. 

Proposed § 53.1425 would address the 

finality of NRC approvals referenced in 

a COL application and is equivalent to 

§ 52.83(a). Proposed § 53.1431 would 

address the referral of COL applications 

to the ACRS for review and is 

equivalent to § 52.87. Proposed 

§ 53.1434 would address the 

authorization to conduct LWA activities 

and is equivalent to § 52.91. Proposed 

§ 53.1437 would address exemptions, 

departures, and variances and is 

equivalent to § 52.93. Proposed 

§ 53.1440 would address issuance of 

COLs and is equivalent to § 52.97. 

Proposed § 53.1443 would address 

finality of COLs and is equivalent to 

§ 52.98. 

Proposed § 53.1449 would address 

inspection during construction and is 

equivalent to § 52.99. Proposed 

§ 53.1452 would address operation 

under a COL and is equivalent to 

§ 52.103. Paragraph (a) of proposed 

§ 53.1452 would include footnotes to 

provide that, for licensees installing 

fueled manufactured reactors under a 

COL, (1) the COL holder would notify 

the NRC of its scheduled date for 

initiating the physical removal of any 

one of the independent physical 

mechanisms to prevent criticality 

required under § 53.620(d)(1) rather 

than its scheduled date for the initial 

loading of fuel, and (2) the NRC would 

time its publication of the notice of 

intended operation based on the COL 

holder’s schedule for initiating the 

physical removal of any one of the 

independent physical mechanisms to 

prevent criticality required under 

§ 53.620(d)(1) rather than the COL 

holder’s scheduled date for the initial 

loading of fuel. These footnotes are 

consistent with the provisions of 

proposed § 53.620(d)(1)(iv), which 

would state that, upon initiating the 

physical removal of any one of the 

independent physical mechanisms to 

prevent criticality in the manufactured 

reactor’s place of operation, the fueled 

manufactured reactor has commenced 

operation. For reactors without the 

independent physical mechanisms to 

preclude criticality under proposed 

§ 53.620(d)(1), operation begins with 

initial fuel load. In both cases, removal 

of the physical features to prevent 

criticality (for reactors with such 

features) and initial fuel load (for 

reactors without such features) put a 

fully constructed utilization facility in a 

position to sustain a nuclear chain 

reaction, and in both cases, the 

utilization facility cannot sustain a 

nuclear chain reaction (for lack of 

sufficient reactivity) until the action 

takes place. Therefore, the NRC 

proposes that initiating the physical 

removal of any one of the independent 

physical mechanisms to prevent 

criticality is the best analogue to initial 

loading of fuel for reactors without such 

features. 

The proposed footnote in § 53.1452(a) 

regarding timing of the notice of 

intended operation for fueled 

manufactured reactors with 

independent physical mechanisms to 

prevent criticality also addresses the 

requirements of Section 189a.(1)(B)(i) of 

the Act. This section requires, in part, 

that ‘‘[n]ot less than 180 days before the 

date scheduled for initial loading of fuel 

into a plant by a licensee that has been 

issued a combined construction permit 

and operating license under section 

185b., the Commission shall publish in 

the Federal Register notice of intended 

operation.’’ That section further requires 

that this notice provide a 60-day period 

in which to request a hearing ‘‘on 

whether the facility as constructed 

complies, or on completion will 

comply, with the acceptance criteria of 

the license.’’ In the case where a fueled 

manufactured reactor arrives at the site 

where it is to be operated by a COL 

holder, the manufacturer would have 

loaded fuel at the factory under its part 

70 license. Therefore, at the site of 

operation, there would not be ‘‘initial 

loading of fuel into a plant by a licensee 

that has been issued a combined 

construction permit and operating 

license’’ (emphasis added). Under a 

literal reading of the entry condition in 

Act Section 189a.(1)(B)(i), this situation 

would not trigger its requirements. 

However, the purpose of the provision 

is to offer the hearing opportunity at 

least 180 days prior to when the fuel is 

loaded and ready for use at its 

authorized location. It would be 

contrary to that purpose if, in this 

situation, the Commission did not 

publish the notice of intended operation 

and opportunity for the public to 

request a hearing on conformance with 

the acceptance criteria in the COL for 

the site of operation. To fulfill the 

underlying purpose of the law, the NRC 

proposes to time the notice of intended 

operation based on the COL holder’s 

schedule for initiating the physical 

removal of any one of the independent 

physical mechanisms to prevent 

criticality required under § 53.620(d)(1). 

This action by the COL holder would be 

the best analogue to initial fuel load by 

the COL holder for the reasons stated 

previously. This analogue is adopted in 

other sections of the proposed part 53 

and related sections in parts 50 and 73 

that use initial fuel loading to identify 


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a transition point for the applicability of 

regulatory requirements. To address the 

possible loading of fuel into a 

manufactured reactor for subsequent 

transport to and use at a commercial 

nuclear plant, multiple sections that 

determine the applicability of 

regulations have been drafted or revised 

to allow for either initial fuel load or 

initiating the physical removal of any 

one of the independent physical 

mechanisms to prevent criticality 

required under § 53.620(d)(1) for a 

fueled manufactured reactor to 

determine the applicability of the 

requirement, as appropriate. 

Proposed § 53.1455 would address 

duration of COL and is equivalent to 

§ 52.104. Proposed § 53.1456 would 

address the transfer of a COL and is 

equivalent to § 52.105. Proposed 

§ 53.1458 would address application for 

renewal and is equivalent to § 52.107. 

Proposed § 53.1461 would address 

continuation of COL and is equivalent 

to § 52.109. 

Proposed § 53.1470 would address 

standardization of commercial nuclear 

plant designs and licenses to construct 

and operate commercial power reactors 

of identical design at multiple sites and 

is equivalent to appendix N of part 52. 

This section would set out the particular 

requirements and provisions applicable 

to situations in which applications for 

CPs and subsequent OLs, or COLs, 

under this part are filed by one or more 

applicants for licenses to construct and 

operate nuclear power reactors of 

identical design (‘‘common design’’) to 

be located at multiple sites. Additional 

information related to this proposed 

section is provided in the final rule to 

revise part 52 (72 FR 49352; August 28, 

2007). 

Subpart I—Maintaining and Revising 

Licensing-Basis Information 

Part 53 would establish requirements 

for the maintenance of licensing-basis 

information in subpart I. 

Section 53.1500 would describe the 

purpose of the subpart in terms of the 

definition of licensing-basis information 

in subpart A. Subpart I would be closely 

tied to the requirements in subpart H, 

which would provide the requirements 

for contents of applications for the 

various types of licenses issued under 

part 53. Subpart I would generally be 

organized into sections dealing with: (1) 

licensing-basis information that 

licensees are not authorized to change 

without NRC approval (e.g., licenses, 

regulations); and (2) licensing-basis 

documents that licensees may change 

provided specified criteria are satisfied 

(e.g., FSAR, program descriptions). The 

subpart would also capture certain 

general conditions on licenses and 

changes to the licenses related to the 

transfer and termination of licenses. 

Section 53.1502 would define specific 

terms and conditions of licenses. These 

terms and conditions would be 

equivalent to the regulations in: (1) 

§ 50.54(h) stating that each license is 

subject to the provisions of the Act and 

requirements issued by the Commission; 

(2) § 50.54(s) stating the actions the 

Commission would take if it makes a 

finding that there is not reasonable 

assurance that adequate protective 

measures can and will be taken in the 

event of a radiological emergency; (3) 

§ 50.54(aa) stating that each license is 

subject to the specified sections of the 

Federal Water Pollution Control Act; 

and (4) § 50.54(dd) stating that a holder 

of an OL or COL may take reasonable 

actions that depart from the license in 

a national security emergency. 

Section 53.1505(a) would serve as an 

introduction to and overview of the 

sections that follow on changes to 

licensing-basis information requiring 

prior NRC approval, namely the 

elements of licensing-basis information 

defined by licenses, orders, and 

regulations. The related sections within 

these subparts would primarily deal 

with the process of how a licensee 

requests and the NRC issues an 

amendment to a license or issues an 

order that modifies a license. Another 

important element of licensing-basis 

information that a part 53 licensee 

would not be able to change or deviate 

from without NRC approval would be 

the NRC regulations themselves. Section 

53.1505(b) would refer to § 53.080 in 

subpart A that would provide the 

criteria for a licensee or other party to 

satisfy when requesting an exemption 

from NRC regulations. 

Section 53.1510 would be equivalent 

to § 50.90 and would require that a 

licensee submit an application to 

request an amendment to a license. The 

required assessments that would be 

included within an application to 

amend a license under part 53 would 

need to address the safety criteria and 

analysis requirements of subparts B and 

C. As with parts 50 and 52, licensees 

would be required to include in their 

applications to amend a license an 

analysis of whether the amendment 

involves no significant hazards 

consideration using the standards in 

§ 53.1520, which would be equivalent to 

the standards in § 50.92. Although this 

rulemaking provided an opportunity to 

revise the terminology related to no 

significant hazards consideration 

determinations, which dates to the early 

1960s when applications were 

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 § 50.91 for the NRC’s 

processes related to applications to 

amend an OL or COL. Section 50.91(b) 

stipulates that the Commission will 

make available to the licensee the name 

of the appropriate State official 

designated to receive such amendments. 

While the Commission intends to 

continue following this practice, the 

Commission has not included this 

administrative matter in proposed part 

53. Proposed § 53.1515(b)(3) contains 

some modifications compared to 

§ 50.91(b)(3) for clarity; these revisions 

are not intended to revise the substance 

of the provisions in part 53 compared to 

part 50. 

Section 53.1520 would be based on 

§ 50.92. The section would continue to 

use the criteria in § 50.92 for 

determining that a proposed 

amendment involves no significant 

hazards consideration. Although more 

specific terms such as event sequence 

are used throughout part 53, § 53.1520 

would use the term ‘‘accident’’ to 

maintain consistency with the long 

history of making no significant hazards 

consideration determinations under part 

50. 

Section 53.1525 would provide 

requirements for holders of an OL or 

COL requesting to revise information 

from a DC rule that was referenced in 

the initial license application and 

included in or incorporated by reference 

into the facility FSAR. In keeping with 

the current requirements in part 52, the 

portion of the part 53 facility licensingbasis information obtained from the 

certified design would be divided into 

two categories. The most significant 

design information and the ITAAC 

would be certified by rule and 

designated as ‘‘certification 

information.’’ The remaining 

information, which makes up the 

majority of the design information 

approved as part of the DC, would not 

be certified by rule and is not 

considered ‘‘certification information.’’ 

Part 52 refers to these categories of 

information as Tier 1 and Tier 2 

information, respectively, and refers to 

a change made to that information on a 

plant-specific basis as a departure. 

Under part 52, a departure from Tier 1 

information requires an exemption and, 


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for information incorporated into the 

license, a license amendment. 

Part 53 would dispense with the Tier 

1 and Tier 2 terminology. Rather, 

§ 53.1525 would use the term 

‘‘certification information’’ in place of 

Tier 1, and a plant-specific departure 

from the certification information would 

require both a request for an exemption 

from the associated DC rule and, for 

information such as ITAAC 

incorporated into the license, a license 

amendment. However, as would be 

provided in § 53.1525(c), a plantspecific departure from the information 

approved by the NRC as part of the DC 

rule but which is not certification 

information (i.e., Tier 2 information 

under part 52) would be assessed using 

the process and criteria defined in 

§ 53.1550 for changes to a FSAR. An 

applicant or licensee would need to 

identify such a change as a departure 

from the referenced standard design in 

the updated FSAR. The process for 

making a generic change to a certified 

design would be described in the 

associated section in subpart H. 

Section 53.1530 would not allow the 

holder of an ML or the holder of a COL 

using a manufactured reactor to make 

changes to the design of the 

manufactured reactor without 

requesting a license amendment from 

the NRC. This section would provide 

the equivalent requirements as those in 

§§ 52.98 and 52.171. 

Section 53.1535 would establish 

requirements for license amendments 

during construction. The section would 

provide the equivalent options and 

requirements for the holders of a CP as 

those in § 50.35(b). The regulations 

would allow but do not require the 

holder of a CP or LWA to request an 

amendment under § 53.1510 if the 

licensee desires to obtain NRC approval 

of a specific design feature or 

specification. The requirements for 

obtaining an amendment to a COL to 

address changes during construction 

would also be provided in § 53.1535. 

The proposed process would differ from 

the current requirements in part 52 by 

adopting a requirement that would 

explicitly support a change process like 

that described in RG 1.237, ‘‘Guidance 

for Changes During Construction for 

New Nuclear Power Plants Being 

Constructed Under a Combined License 

Referencing a Certified Design Under 10 

CFR part 52.’’ 

The proposed regulation would allow 

the holder of a COL to proceed at its 

own risk in making a change during the 

construction process and would require 

that licensee to submit a license 

amendment request no later than 45 

days from the date the licensee begins 

to implement the change or departure 

requiring NRC approval. 

Section 53.1540 would serve as an 

introduction to the sections that follow 

on changes to licensing-basis 

information that are primarily under the 

control of a licensee but for which 

evaluations are made to determine if a 

submittal to the NRC requesting 

approval would be required. The section 

would also include definitions that 

would be applicable when using the 

processes in §§ 53.1545 through 

53.1565. The definitions would be 

largely equivalent to those in § 50.59(a) 

but include some revision to reflect the 

structure and terminology in other 

subparts in part 53. For example, the 

definition of ‘‘Change’’ in § 53.1540(b) 

would address a ‘‘design feature or 

related functional design criteria’’ rather 

than a ‘‘design function,’’ because the 

former are defined terms in part 53. 

Similarly, in § 53.1540(b), the phrase 

‘‘design basis’’ from § 50.59(a)(2) would 

be replaced with functional design 

criteria for SR SSCs. 

Section 53.1545 would provide the 

proposed requirements for updating of 

FSARs. While the process-related 

requirements proposed under § 53.1545 

would be largely the same as those in 

§ 50.71, the specifics of information to 

be updated would differ due to the role 

of PRA in satisfying the requirements in 

subparts B and C. Additionally, the use 

of the risk-informed approach in subpart 

C would result in some but not all PRA 

information being in the FSAR or 

another licensing basis document and 

therefore a separate PRA update 

requirement similar to § 50.71(h) is not 

included in proposed subpart I. 

Proposed § 53.1239(a)(18) in subpart 

H and the related references to this 

proposed requirement for the holders of 

OLs and COLs would require a 

description of the PRA required by 

§ 53.450(a) and its results to be included 

in FSARs. However, guidance 

documents are planned to clarify the 

division of PRA-related information that 

would need to be in the FSAR, in other 

possible licensing basis documents, and 

controlled as plant records subject to 

inspections and audits. At a minimum, 

the information from the PRA that 

would be needed to show compliance 

with subpart C would be included in the 

FSAR (e.g., PRA summary and 

analytical results for LBEs). The 

submittal of voluminous PRA 

information was initially required under 

part 52, but that proved to be 

impractical and was revised in the 2007 

revision of part 52. Guidance is being 

developed to ensure sufficient 

information is submitted to the NRC to 

support the licensing process and the 

NRC’s regulatory findings under part 53 

or similar applications using the LMP 

under parts 50 or 52. 

The NRC has posed a question in 

section VI, ‘‘Specific Requests for 

Comments,’’ of this document that asks 

about the appropriate level of detail for 

PRA-related information in an FSAR 

and whether other licensing basis 

documents might be more appropriate 

to both provide information to the NRC 

and ensure the PRA is maintained and 

updated as proposed in subpart C. The 

program document would provide more 

detail than the summaries in the FSAR 

but still be a much-condensed source of 

information in comparison to the 

documentation of the PRA. 

Section 53.1545(a)(3) and (4) would 

be based on the inclusion of at least a 

summary of PRA results and the related 

margins to safety criteria in the FSAR 

and would require updates to that 

information. The routine reporting of 

these margins would also inform 

application of the criteria for allowing 

changes without an amendment in the 

following section (§ 53.1550) in subpart 

I. 

Section 53.1550 would establish 

requirements for evaluating changes to a 

facility as described in its FSAR. This 

proposed section would provide the 

equivalent of the requirements in 

§ 50.59 for evaluating changes to an 

FSAR (as updated) and determining if a 

license amendment is required to 

implement a change to a facility or 

procedures. The evaluation criteria 

proposed in § 53.1550 would reflect the 

role of the PRA in the safety analyses 

under part 53 and would include 

several measures related to the changes 

in plant risk resulting from a change in 

the plant design or plant procedures. 

Examples would include criteria that 

rely on the identification of risksignificant event sequences in 

accordance with the analysis 

requirements of § 53.450; exceeding the 

LBE evaluation criteria as defined in 

§ 53.450; the consideration of potential 

reductions in margin between the 

estimated comprehensive risk metrics 

and associated risk performance 

objectives in the safety criteria in 

§ 53.220; changes to the safety 

classification of SSCs; and consideration 

of reductions in defense in depth. 

Section 53.1550 would include a 

criterion related to a departure from a 

method of evaluation used in the safety 

analyses. The NRC has not yet 

developed draft guidance for use in 

applying proposed § 53.1550 but 

anticipates that the NRC and 

stakeholders will assess the potential 

need for such guidance and that such 

guidance would, if needed, be 


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developed as part of ongoing or future 

activities. 

Section 53.1550 would include 

certain concepts taken from existing 

guidance for § 50.59 in the proposed 

criteria related to DBAs. Specifically, 

criterion (iv) for changes made to a 

method of evaluation of DBAs under 

§ 53.450(f) would be equivalent to a 

change in a method of evaluation under 

§ 50.59, and criterion (viii) on assessing 

if a change creates a possibility for an 

accident of a different type than 

previously analyzed in the FSAR would 

be similar to the § 50.59 criterion (v). 

Guidance documents will be prepared 

to address the content of applications 

for PRA-related information under 

proposed part 53, and this guidance will 

also influence how potential changes in 

the evaluation of LBEs other than DBAs 

analyzed under § 53.450(e) are 

evaluated and reported under the 

proposed criterion (iv). 

Section 53.1550(a)(2)(x) would 

require evaluating plant changes to 

ensure they would not prevent 

satisfying the design requirements in 

§ 53.440(j) related to the impact of a 

large commercial aircraft. The inclusion 

of a proposed requirement under 

§ 53.1550 related to design features for 

protecting against aircraft impact would 

reflect the proposed design requirement 

in subpart C and related proposed 

requirements in subpart H to address 

the proposed design requirement in 

FSARs. 

Sections 53.1560 through 53.1565 in 

subpart I would define the processes for 

a licensee to evaluate changes to the 

program documents included in the 

licensing-basis information submitted to 

the NRC and to modify such programs 

without NRC prior approval. 

Section 53.1560 would include the 

proposed requirements for updating 

program documents included in 

licensing-basis information and would 

provide the equivalent of FSAR updates 

for key program documents. The 

proposed requirements in these sections 

would provide a uniform approach for 

updating program documents, which 

correspond to the programs required 

under subpart F. 

The proposed § 53.1565 would 

provide a process for licensees to make 

changes to program documents included 

in licensing-basis information without 

obtaining prior NRC approval. The 

proposed requirements would include 

several generic criteria that, if not 

satisfied, would prompt the need for 

NRC approval of a change to a program 

document. These generic criteria would 

include whether a change would 

comply with TS and NRC regulations. 

Another proposed criterion for 

evaluating changes to program 

documents would be conforming with 

program-specific requirements, 

including NRC-approved program 

documents with more specific criteria 

for a particular program, regulations, 

administrative controls sections of TS, 

and NRC-approved program documents. 

Proposed § 53.1565(d) would include 

specific criteria for evaluating changes 

to several program documents that have 

well established change processes and 

guidance for licensees under parts 50 

and 52. The program documents 

specifically addressed in the proposed 

section would include quality assurance 

programs that would be equivalent to 

§ 50.54(a), an emergency preparedness 

program that would be equivalent to 

§ 50.54(q), and the security program that 

would be equivalent to § 50.54(p). 

The proposed § 53.1570 would 

establish requirements for the transfer of 

commercial nuclear plant licenses by 

providing the equivalent requirements 

of § 50.80 for the possible transfer of an 

ESP, CP, OL, or COL. Likewise, the 

proposed § 53.1575 would establish 

requirements for the termination of an 

OL or COL by providing the equivalent 

requirements of § 50.82. Other proposed 

requirements related to 

decommissioning and license 

termination would be included in 

subpart G. 

Section 53.1580 would establish 

requirements for information requests 

the NRC could send to the various types 

of licensees and would provide 

requirements that would be equivalent 

to requirements in § 50.54(f). The 

proposed § 53.1585 would provide the 

requirements that would be equivalent 

to requirements in § 50.100 to address 

revocation, suspension, modification of 

licenses, and approvals for cause. 

Section 53.1590 would propose to 

address backfitting requirements by 

providing requirements that would be 

equivalent to those in § 50.109. 

Proposed § 53.1595 would address 

license renewals under part 53 with 

simple statements that licenses may be 

renewed. This section would be 

expanded through future rulemakings to 

more fully describe or reference the 

processes related to requesting and 

processing applications to renew ESPs, 

OLs, and COLs issued under part 53 (if 

finalized). 

Subpart J—Reporting and Other 

Administrative Requirements 

Part 53 would address various 

reporting and administrative 

requirements in subpart J. 

Section 53.1600 would explain the 

organization of the various sections 

within the subpart related to providing 

unfettered access to NRC inspectors; 

maintaining certain records and 

reporting specified events or conditions; 

demonstrating compliance with 

financial qualification requirements and 

providing specified financial reports; 

and maintaining financial protections to 

address potential accidents. 

Section 53.1610 would establish 

requirements for the provision of 

facilities and unfettered access for 

inspections. These requirements would 

be equivalent to § 50.70 with only minor 

changes proposed to provide additional 

flexibilities and address possible 

differences related to reactors licensed 

under part 53 and the possibility that 

some commercial nuclear plants may 

not be assigned resident inspectors. 

Section 53.1620 would provide for 

maintenance of records and the making 

of various reports to the NRC. These 

requirements would be largely 

equivalent to § 50.71. This section is not 

intended to reflect all provisions in 

§ 50.71; several important requirements 

in § 50.71 would be captured in other 

sections of part 53. For example, 

§ 53.1545 within subpart I would 

provide requirements that would be 

equivalent to § 50.71(e), updating 

FSARs, and § 53.1680, ‘‘Annual 

financial reports,’’ would provide the 

equivalent of § 50.71(b), which covers 

financial reports. A reporting 

requirement related to completion of 

power ascension testing would be added 

to § 53.1620 to support the assessment 

of annual fees under 10 CFR part 171, 

‘‘Annual Fees for Reactor Licenses and 

Materials Licenses, Including Holders of 

Certificates of Compliance, 

Registrations, and Quality Assurance 

Program Approvals and Government 

Agencies Licensed by the NRC,’’ which 

normally commence upon completion 

of those testing activities. 

Section 53.1630 would establish 

requirements for immediate notification 

requirements for operating commercial 

nuclear plants. These requirements 

would be equivalent to § 50.72 with 

minor changes proposed to make the 

reporting criteria technology inclusive. 

In addition, a new version of NRC Form 

361 (NRC Form 361S) would be created 

for use by part 53 licensees, but without 

LWR-specific terminology to ensure 

technology inclusiveness. A separate 

rulemaking activity, ‘‘Reporting 

Requirements for Nonemergency Events 

at Nuclear Power Plants,’’ has been 

initiated to consider possible changes to 

the requirements in § 50.72. At a future 

date, the NRC may consider reconciling 

future changes to § 50.72 with the 

requirements proposed in part 53, 

which have been taken or derived from 

the current reporting requirements. 


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3A ‘‘merchant plant’’ is a plant licensed to a nonrate-regulated entity (e.g., a nonutility) that engages 

in the business of production, manufacturing, 

generating, buying, aggregating, marketing, or 

brokering electricity for sale at wholesale or for 

retail sale to the public. 

4The 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. 

Section 53.1640 would address the 

licensee event report system. These 

requirements would be equivalent to 

§ 50.73 with minor changes proposed to 

make the requirements inclusive of 

various reactor technologies and to 

reflect appropriate internal references to 

other sections in part 53. In addition, 

NRC Forms 366, 366A, and 366B would 

be revised to include corresponding 

check boxes for part 53 licensees. 

Section 53.1645 would require 

periodic reporting of the quantity of 

radionuclides released to unrestricted 

areas in liquid and gaseous effluents, 

doses to members of the public, and the 

results of environmental monitoring. 

These reporting requirements in the 

proposed part 53 would be largely 

equivalent to those in the TSs required 

by § 50.36a, ‘‘Technical specifications 

on effluents from nuclear power 

reactors.’’ The only difference would be 

that a § 50.36a requirement to 

specifically address conditions where 

the dose to the maximally exposed 

individual could be significantly above 

design objectives would refer to a design 

objective of 10 mrem/year total effective 

dose equivalent, instead of referring to 

the design objectives in appendix I to 

part 50. The proposed section would 

also include an equivalent to the 

reporting requirement in section IV of 

appendix I to part 50 if the radiation 

exposure to a member of the public in 

any calendar quarter exceeds one-half of 

the annual ALARA design objective. 

Section 53.1650 would include a 

reporting requirement to support 

safeguards agreements between the 

United States and the International 

Atomic Energy Agency (IAEA) and 

would be equivalent to § 50.78. 

Section 53.1660 through 53.1700 

would address financial requirements 

and would be largely similar to existing 

regulations in parts 50 and 52. Section 

53.1670 would be entitled ‘‘Financial 

qualifications’’ and would require 

applicants other than electric utilities to 

possess or have reasonable assurance of 

obtaining funds for the activities for 

which the license is being sought. The 

NRC is seeking feedback on these 

sections and their ramifications for 

merchant plants 3 in section VI, 

‘‘Specific Requests for Comments,’’ of 

this document. The remaining financial 

reports in part 53 would be equivalent 

to § 50.71(b) for annual financial 

reports, § 50.76 for a change of status, 

§ 50.54(cc) for the filing of a petition for 

bankruptcy, and § 50.81 for creditor 

regulations. 

Sections 53.1710 through 53.1730 

would address financial protection 

requirements. Section 53.1720 would 

require insurance to stabilize and 

decontaminate a plant following an 

accident. These requirements would be 

taken from § 50.54(w) with the only 

notable change being the addition of a 

provision allowing plant-specific 

estimates of costs to stabilize and 

decontaminate a plant as an alternative 

to the $1.06 billion minimum coverage 

in § 50.54(w). Section 53.1730 is 

equivalent to § 50.57(a)(5) and would 

refer to the requirements in 10 CFR part 

140, ‘‘Financial Protection 

Requirements and Indemnity,’’ related 

to financial protection requirements and 

indemnity agreements, including the 

financial protection requirements of the 

Price-Anderson Act. 

Subpart M—Enforcement 

Subpart M would contain two 

provisions, § 53.9000 and § 53.9010, 

which are analogous to provisions 

contained in other parts of 10 CFR 

Chapter I imposing requirements on 

regulated entities. Section 53.9000 

would provide notice of the 

Commission’s authority under the Act 

to obtain injunctions or other court 

orders for the enumerated violations. 

Paragraph (a) of § 53.9010 would 

provide notice to all persons and 

entities subject to part 53 that they are 

subject to criminal sanctions for willful 

violations, attempted violations, or 

conspiracy to violate certain regulations 

under part 53. Criminal sanctions would 

not apply to the regulations listed in 

paragraph (b). The regulations for which 

criminal penalties would apply are 

limited to those that establish either a 

regulatory obligation or prohibition. 

V. Changes to Other Parts of 10 CFR 

Chapter I 

10 CFR Part 26 

A. Introduction 

The NRC is proposing a technologyinclusive, risk-informed, and 

performance-based approach for the 

application of drug and alcohol testing 

and fatigue management requirements 

for facilities licensed under part 53. The 

proposed requirements applicable to 

these applicants, licensees, and other 

entities would be commensurate with 

the radiological consequences presented 

by the applicants’ facilities and the 

operation of these facilities.4 The 

proposed FFD framework would consist 

of a two-tiered graded approach similar 

to that currently in part 26 and an 

optional third tier for part 53 

commercial nuclear plants that perform 

an analysis that demonstrates the 

facility and its operation would satisfy 

the criterion in proposed § 26.603(c), 

which refers to § 53.860(a). This 

proposed FFD framework would be 

established in subpart M, ‘‘Fitness for 

Duty Programs for Facilities Licensed 

Under Part 53,’’ of part 26. 

The NRC used operating experience to 

provide regulatory flexibility in the 

proposed subpart M of part 26 

framework to help support a licensee’s 

or other entity’s response to changes in 

societal drug use, drug testing 

technologies and processes, and FFD 

program performance. The flexibility 

would also help in FFD program 

implementation because of the wide 

variety of staff sizes anticipated at 

commercial nuclear plants licensed 

under part 53 and the geographically 

remote locations in which commercial 

nuclear plants may be sited. 

The proposed first-tier FFD program 

requirements would apply to part 53 

licensees and other entities of 

commercial nuclear plants under 

construction who satisfy the criterion in 

§ 26.603(c) but elect not to implement 

proposed § 26.604, ‘‘FFD program 

requirements for facilities that satisfy 

the § 26.603(c) criterion,’’ or who do not 

satisfy the criterion in § 26.603(c), and 

to holders of MLs who are assembling 

or testing manufactured reactors. These 

requirements would be provided in 

proposed § 26.605(a) and would be 

essentially equivalent to those 

requirements in subpart K, ‘‘FFD 

Program for Construction,’’ of part 26 as 

supplemented by select requirements 

from subparts E, ‘‘Collecting Specimens 

for Testing,’’ and I, ‘‘Managing Fatigue,’’ 

of part 26, and the requirements in 

subparts A, ‘‘Administrative 

Provisions,’’ and O, ‘‘Inspection, 

Violations, and Penalties,’’ of part 26. 

The first-tier requirements would 

involve policies, procedures, behavioral 

observation, fatigue management, drug 

and alcohol testing, determinations of 

fitness, appeals, training, sanctions, 

auditing, change control, performance 

monitoring, recordkeeping, and 

reporting. These requirements would 

help deter individuals subject to this 

section from illicit drug and/or alcohol 

use and from being impaired from any 

cause including fatigue. These proposed 

requirements would also help licensees 


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5Controlled access area and vital area are defined 

in § 73.2, ‘‘Definitions.’’ 

and other entities identify individuals 

as users of impairing substances and 

demonstrate compliance with § 26.23, 

‘‘Performance objectives.’’ 

The proposed second tier would 

include all the proposed first-tier 

requirements, plus the more 

comprehensive set of FFD program 

requirements in current subparts C, 

‘‘Granting and Maintaining 

Authorization,’’ D, ‘‘Management 

Actions and Sanctions to be Imposed,’’ 

H, ‘‘Determining Fitness-for-Duty Policy 

Violations and Determining Fitness,’’ 

and N, ‘‘Recordkeeping and Reporting 

Requirements,’’ of part 26. These 

requirements would be provided in 

proposed § 26.605(b) and would be 

applicable to licensees and other 

entities satisfying the § 26.603(c) 

criterion, at their discretion. These 

requirements would also apply to 

licensees or other entities not satisfying 

the § 26.603(c) criterion that implement 

an FFD program under subpart M of part 

26, before the loading of fuel onsite into 

a reactor vessel; before receiving a 

manufactured reactor; or before 

operating, testing, performing 

maintenance of, or directing the 

maintenance or surveillance of securityrelated equipment or equipment that a 

risk-informed evaluation process has 

shown to be significant to public health 

and safety. 

The second-tier requirements are 

based on the additional risk presented 

by nuclear reactor assembly, testing, 

fueling, and operation and the necessity 

for human actions in certain event 

sequences. The inclusion of the current 

part 26 requirements would align 

proposed part 53 FFD and AA program 

requirements with the current FFD and 

AA programs required for facilities 

licensed under parts 50 and 52. This 

approach would ensure effective and 

consistent AA and FFD program 

implementation across the commercial 

nuclear power industry, thereby 

ensuring uniform requirements for 

individuals who may perform roles and 

responsibilities for multiple facilities 

regardless of facility licensure. 

Proposed § 26.604 would offer an 

alternate option for an applicant 

implementing an FFD program under 

subpart M of part 26. If the applicant 

demonstrates that the criterion in 

proposed § 26.603(c) is met, then the 

applicant (and the subsequent licensee 

or other entity) must still implement an 

FFD program described in subpart M of 

part 26; however, drug and alcohol 

testing would not be required unless 

FFD performance declines or the 

applicant, licensee, or other entity elects 

to implement drug and alcohol testing. 

The proposed § 26.604 requirements are 

equivalent to those proposed in 

§ 26.605(a) except for required drug and 

alcohol testing. This proposed 

framework would focus on the human 

performance of individuals while they 

are performing those duties and 

responsibilities that make them subject 

to the FFD program. This performance 

would be verified through behavioral 

observation, evaluation of any FFD 

concerns, performance monitoring, 

fatigue management, and 

determinations of fitness. Applicants 

that do not satisfy the criterion in 

proposed § 26.603(c), or elect not to 

perform the analysis required to 

demonstrate that the criterion in 

§ 26.603(c) is met, would be subject to 

an FFD program described in § 26.605, 

‘‘FFD program requirements for 

facilities that do not implement 

§ 26.604,’’ or an FFD program that 

implements all part 26 requirements, 

except for those requirements in 

subparts K and M of part 26. 

In establishing the minimum FFD 

program requirements in § 26.604, the 

NRC reviewed current advanced reactor 

designs against that of a non-power 

production or utilization facility (NPUF) 

that is not required to implement an 

FFD program for those individuals who 

have unescorted access to the controlled 

access area (and vital area for some 

facilities), including NRC-licensed 

operators.5 This review was performed 

because commercial nuclear plants 

licensed under part 53 could be 

designed with similar power levels and 

radiological consequences as the 

currently licensed NPUFs. From this 

review, three principal considerations 

supported the minimum set of 

requirements for the § 26.604 FFD 

program. 

First, the radiological consequences 

presented by a part 53 licensed facility 

and its operation that satisfy the 

criterion in § 26.603(c) may present a 

greater potential radiological 

consequence to workers and the public 

in the vicinity of the facility than does 

an NPUF. Second, the operating 

characteristics of a part 53 licensed 

facility are unlike that of an NPUF 

because there may be a higher reliance 

on individuals at the part 53 site to 

safely and competently operate, 

maintain, surveil, and secure SSCs that 

may not be required at an NPUF, such 

as systems that provide secondary heat 

transfer, reactor coolant flow, pressure 

control, and at-power core refueling. 

Differences in operating characteristics 

could include, for example: long-term, 

full power operation with automated 

reactivity control systems for loadfollowing; active and passive safety and 

security systems; innovative non-lightwater heat transfer systems; and energy 

storage and hazardous chemical 

systems. The individuals at part 53 

facilities may also be required to 

communicate to individuals both onsite 

and offsite, such as electrical load 

dispatchers, any conditions adverse to 

safety, security, or quality. Third, part 

53 licensed facilities may be sited in 

geographically remote locations that 

may not have a physically available 

administrative or corporate support 

team to provide face-to-face oversight, 

engineering expertise, and maintenance 

support like that at NPUFs. This places 

a higher reliance on those individuals 

required at a part 53 facility being fit for 

duty and trustworthy and reliable 

because a replacement individual may 

not be readily available. 

The NRC proposes to exclude drug 

and alcohol testing from the proposed 

§ 26.604 framework for five reasons: (1) 

the § 26.23 performance objectives can 

be met through effective 

implementation of the defense-in-depth 

regulatory framework established by 

behavioral observation, reporting of 

legal actions, the proposed performance 

monitoring and review program (PMRP), 

FFD training, and requirements from the 

physical protection, AA, cyber 

protection, and licensed operator 

programs; (2) the PMRP would require 

the licensee or other entity to monitor 

its FFD program performance (both 

qualitatively and quantitatively) against 

its historical site performance, fleetlevel performance, if applicable, and 

industry performance. The licensee or 

other entity would be required to 

implement corrective actions if site FFD 

performance meets a licensee- or other 

entity-established threshold or to 

resolve a finding resulting from a 

qualitative review or audit in a manner 

that restores performance and corrects 

root causes, contributing causes, or 

both; (3) the requirements in proposed 

§ 26.609, ‘‘Behavioral observation,’’ are 

more robust than those in § 26.407, 

‘‘Behavioral observation,’’ of subpart K 

of part 26 and are proposed to 

synchronize with and reinforce the AA 

behavioral observation requirements in 

§ 73.56, ‘‘Personnel access authorization 

requirements for nuclear power plants,’’ 

or the proposed requirements under 

§ 73.120, ‘‘Access authorization program 

for commercial nuclear plants’’; (4) a 

part 53 commercial nuclear plant that 

satisfies the § 26.603(c) criterion will be 

designed, operated, and secured with a 

radiological risk profile that is lower 

than that described in § 53.860(a)(2) and 


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6See e.g., Irene W#r<, Ragnar Rosness, and Stine 

Skaufel Kilska, ‘‘Human performance and safety in 

Arctic environments,’’ SINTEF (2018). 

7See 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). 

8See e.g., W#r<, Rosness, and Kilska (arguing 

that groupthink leads teams to ‘‘develop shared 

rationalizations that bolster a proposed choice, 

rather than examining alternative options and 

identifying the risks associated with the proposed 

choice’’). See also David Hofmann and Adam 

Stetzer, ‘‘A Cross-Level Investigation of Factors 

Influencing Unsafe Behaviors and Accidents,’’ 

Personnel Psychology, Vol. 49 (1996) (finding that 

in a study of fatal accidents involving offshore oil 

rigs, in the absence of standard operating 

procedures, workers ‘‘equated normal work 

methods (i.e. what everyone else does) with safe 

and/or ideal work methods,’’ revealing that the 

groupthink phenomena will further cement modes 

of work that do not reflect safety protocols in small 

groups that lack strong norms around workplace 

safety and tacitly reward short-cuts that prioritize 

efficiency over safety). 

9Mannion and Thompson, ‘‘Systematic biases in 

group decision-making: implications for patient 

perhaps will approach the radiological 

risk profile of an NPUF (which does not 

implement an FFD program); and (5) the 

NRC is aware that a part 53 commercial 

nuclear plant could be designed and 

constructed in such a manner to reduce 

reliance on an onsite security force to 

protect SSCs, NRC-licensed materials, 

and sensitive information, with 

enhanced capabilities for the detection, 

assessment, and delay of a DBT 

adversary. 

Regarding fatigue management 

requirements, work hour controls would 

be required for personnel at utilization 

and manufacturing facilities in 

accordance with the existing scoping 

criteria in § 26.4, ‘‘FFD program 

applicability to categories of 

individuals,’’ as revised in this 

proposed rule. The amended § 26.4 also 

would be used to determine whether an 

individual would be subject to drug and 

alcohol testing. The applicability of 

these scoping criteria for certain 

individuals (such as operators and 

maintenance personnel) would be 

determined by the licensee or other 

entity through its risk-informed 

evaluation process performed to assess 

the risk significance of the SSC upon 

which work is being performed or 

directed by the individual. These 

requirements also would be scaled 

based on the potential radiological 

consequences presented by the facility. 

However, fatigue management would be 

applied to all individuals subject to the 

FFD program, similar to FFD program 

implementation by the current fleet of 

commercial nuclear plants because 

fatigue management is a proactive 

requirement designed to help prevent 

on-shift impairment through work hour 

scheduling and time off. The behavioral 

observation program (BOP) would be 

the principal requirement to provide 

reasonable assurance that individuals 

on shift are not mentally or physically 

impaired due to fatigue, which in any 

way could adversely affect their ability 

to safely and competently perform their 

duties. 

The NRC is proposing subpart M of 

part 26 for facilities licensed under part 

53, in lieu of subjecting all part 53 

licensees to the same part 26 

requirements that apply to facilities 

licensed under part 50 or 52, for four 

principal reasons. First, subpart M of 

part 26 would apply FFD requirements 

in a risk-informed manner 

commensurate with the radiological 

consequences presented by facilities 

licensed under part 53. This regulatory 

strategy is consistent with the current 

part 26, which provides a 

comprehensive set of deterministic 

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 § 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 decisionmaking pressures may be high. 

Groupthink could have adverse effects 

on workplace safety culture, as studies 

show that individuals will be more 

hesitant to speak out against practices 

they deem unsafe for fear of deviating 

from group norms.7 Individuals may 

also be unaware of systematic biases in 

the group decision-making process and 

may then be less likely to scrutinize the 

potential risks of the group’s decision or 

sufficiently contemplate alternative 

paths of action.8 Furthermore, the 

literature indicates that groups make 

riskier decisions than individuals acting 

alone due to the diffusion of 

responsibility among group members.9


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safety,’’ International Journal for Quality I Health 

Care, Vol. 26, No. 6 (2014): 606–612. 

This phenomenon, known as ‘‘the risky 

shift,’’ also runs counter to a safety 

culture. Accordingly, ‘‘groupthink’’ and 

‘‘the risky shift’’ may lead to group 

behaviors that render behavioral 

observation less effective. As such, 

alternative approaches to behavior 

observation programs, such as the 

utilization of video-based surveillance 

by individuals separate from the onsite 

work unit, could serve to mitigate 

potential issues associated with 

groupthink. The incorporation of remote 

observation, performed by individuals 

physically separate from the site, could 

help to bring in independent and 

objective perspectives and help to break 

patterns of thought and communication 

that may result in groupthink. 

Even without the influence of small 

group dynamics, there are other 

practical constraints to implementing 

FFD requirements, such as random drug 

and alcohol testing, among small staffs. 

Random testing is less effective when 

applied to small staff sizes because it 

may be easier for staff to communicate 

and predict when individuals will be 

subject to drug and alcohol testing. 

Furthermore, if a facility is sited in a 

remote location, program 

implementation could be challenged by 

the following factors: limited mail 

services to laboratories certified by the 

U.S. Department of Health and Human 

Services (HHS), availability of local 

clinical or medical options for treatment 

and determinations of fitness by an 

MRO or Substance Abuse Expert, and 

use of offsite drug and alcohol 

collection facilities. 

The increased potential for small staff 

sizes to impact FFD policy compliance 

warrants an approach to FFD that 

emphasizes performance over 

prescriptive requirements that may be 

ineffective or infeasible at these 

facilities. Therefore, the NRC proposes 

the subpart M of part 26 framework to 

provide a performance-based approach 

to FFD. For example, proposed 

§ 26.603(d) would use existing part 26 

auditing requirements and the reporting 

requirement in § 26.717, ‘‘Fitness-forduty program performance data,’’ and 

clarify how FFD performance data 

would be used to maintain or improve, 

if necessary, FFD program effectiveness. 

Specifically, § 26.603(d) would require 

each licensee and other entity that elects 

to implement subpart M of part 26 to 

monitor and assess their site-specific 

performance against the preceding 

year’s site performance, the licensee’s 

most recent fleet-level performance, and 

the most recent industry performance. 

Licensees and other entities would use 

these datapoints to develop performance 

measures, which would be qualitative 

descriptions of the specific FFD 

program elements, and threshold values 

for each performance measure that, if 

exceeded, would indicate a performance 

deficiency. Each licensee and other 

entity would compare its site’s current 

performance data against the 

performance measures and, if a 

threshold is exceeded, the licensee or 

other entity would be required to take 

corrective actions to restore 

performance. Also, the NRC proposes a 

change control requirement to allow a 

licensee or other entity to change its 

subpart M of part 26 FFD program while 

ensuring that FFD program effectiveness 

is maintained. 

Lastly, subpart M of part 26 would 

consolidate the applicable FFD 

requirements by placing in one subpart 

all proposed part 26 requirements 

(either new requirements or crossreferences to existing part 26 

requirements) for part 53 licensees and 

other entities. This should help 

licensees and other entities implement 

the requirements because it would 

enable easy cross-reference to similar 

requirements in other subparts that are 

being implemented by non-part 53 

licensees and entities subject to part 26. 

Understanding how other licensees or 

other entities implement similar FFD 

requirements may facilitate the sharing 

of operating experience in program 

implementation. 

The use of innovative technologies 

and a risk-informed performance-based 

framework parallels the considerations 

presented in the Advanced Reactor 

Policy Statement. As stated in the policy 

statement, ‘‘[S]implified systems should 

facilitate operator comprehension, 

reliable system function, and more 

straightforward engineering analysis.’’ 

Furthermore, these same attributes may 

reduce potential radiation exposures, 

help prevent the theft of nuclear 

materials, and use technology and 

design innovations. Should these 

components and systems be designed, 

implemented, and maintained to 

minimize reliance on human actions 

and leverage technology and innovation, 

then the robust and prescriptive FFD 

requirements in, for example, subparts 

B, ‘‘Program Elements,’’ and E of part 26 

could be scaled to the part 53-licensed 

facility and its operation. This strategy 

would be implemented in the subpart M 

of part 26 framework. 

Even though current subpart K of part 

26, provides for FFD requirements 

commensurate with the radiological 

consequences presented by a nuclear 

power plant construction site, proposed 

subpart M of part 26 would not allow 

part 53 licensees and other entities to 

implement the requirements in subpart 

K. The principal reasons are that 

(without significant changes to subpart 

K that would be outside the scope of 

this rulemaking): (1) subpart K does not 

apply to holders of MLs who assemble 

or test a reactor; (2) subpart K only 

applies during construction, whereas 

subpart M would apply during 

construction, operation, and 

decommissioning through 

implementation of the insider 

mitigation program (IMP) required by 

§ 73.55 or proposed § 73.100; (3) subpart 

K does not address training, 

authorization as defined in § 26.5, and 

MRO performance; (4) subpart K does 

not expressly authorize the use of 

innovative drug and alcohol testing 

technologies; (5) subpart K does not 

describe the use of time-dependent 

alcohol limits or special analysis testing 

of dilute urine specimens; and (6) 

subpart K has less rigor in the protection 

of worker rights and sensitive 

information than that proposed in 

subpart M. 

Despite the differences between 

subparts K and M of part 26, the 

requirements in subpart M would be 

essentially equivalent to many in 

subpart K that were implemented by the 

licensees of Vogtle Nuclear Station and 

V.C. Summer Nuclear Station when they 

were constructing four commercial 

nuclear power reactors and NRC 

inspection and operating experience 

evaluation determined that the use of 

subpart K contributed to adequately 

protecting the public health and safety 

and the common defense and security. 

Further, given the risk profile posed by 

facilities licensed under part 53 and the 

proposed additional requirements in 

subpart M of part 26 that were 

developed from operating experience 

and other part 26 subparts (but are not 

included in subpart K of part 26), the 

NRC concludes that if licensees and 

other entities effectively implement the 

proposed requirements in subpart M of 

part 26, then individuals subject to the 

rule should be fit for duty and 

trustworthy and reliable. 

B. Proposed Changes to Part 26, 

Subparts A Through E and I 

Section 26.3(d) is the applicability 

paragraph for contractor/vendors (C/Vs) 

who implement FFD programs or 

program elements, to the extent that the 

licensees and other entities specified in 

§ 26.3(a) through (c) rely on those C/V 

FFD programs or program elements to 

meet the requirements of part 26. 

Section 26.3(d) would be amended to 


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address part 53 licensees and other 

entities in proposed § 26.3(f). 

Proposed § 26.3(f) would place part 53 

licensees or other entities within the 

scope of part 26. For licensees and other 

entities of a part 53 commercial nuclear 

plant, except a holder of an ML, the FFD 

program would be required to be 

implemented no later than the start of 

construction activities. The holder of an 

ML would need to implement its FFD 

program before commencing activities 

that assemble a reactor. 

Current § 26.4 describes FFD program 

applicability to categories of 

individuals. These categories are based 

on the duties, responsibilities, and the 

types of access an individual may 

possess. The NRC proposes to amend 

§ 26.4 to include licensees and other 

entities described in § 26.3(f). The NRC 

expects that not all categories of 

individuals described in current § 26.4 

would be applicable to all part 53 

facilities. The NRC is proposing 

regulatory guidance in DG–5073, 

‘‘Fitness-of-Duty Programs for 

Commercial Nuclear Plants and 

Manufacturing Facilities Licensed 

Under 10 CFR part 53,’’ and DG–5078, 

‘‘Fatigue Management for Nuclear 

Power Plant Personnel at Commercial 

Nuclear Plants Licensed Under 10 CFR 

part 53,’’ to help address program 

applicability to certain individuals. 

Section 26.4(a)(1) and (a)(4) would be 

amended to account for the possibility 

that certain individuals may perform or 

direct the performance of operational 

and maintenance activities from a 

remote facility (for example, a remotecontrol station) for licensees or other 

entities licensed under part 53. 

The framework of the current part 26 

does not account for individuals who 

perform operating and maintenance 

duties at remote facilities. Although 

current § 26.4(a)(1) does not limit the 

operating of applicable SSCs to onsite 

operating, § 26.5 limits the definition of 

‘‘Maintenance,’’ for the purposes of 

§ 26.4(a)(4), to include only ‘‘onsite 

maintenance activities.’’ In the 2008 

part 26 final rule preamble, the NRC 

explained that the work hour 

requirements apply to those individuals 

who perform maintenance activities 

within the licensee’s owner-controlled 

area. Furthermore, regarding the 

direction of applicable operations and 

maintenance activities, current 

§ 26.4(a)(1) and (4) address only 

individuals who perform ‘‘onsite 

direction.’’ 

Under the proposed amendments to 

part 26, the limitation of ‘‘onsite’’ 

activities to those performed within the 

owner-controlled area would still apply 

to facilities licensed under part 50 or 52. 

However, for licensees and other 

entities described in § 26.3(f), the NRC 

would remove the ‘‘onsite’’ limitation to 

include activities performed both within 

the owner-controlled area as well as 

operations and maintenance duties 

performed at remote facilities where 

safety-significant systems and 

components are expected to be operated 

within the design basis of the 

commercial nuclear plant. 

In the 2008 part 26 final rule, the 

purpose of limiting ‘‘directing’’ 

activities to those ‘‘directing’’ activities 

that are conducted onsite was to avoid 

requiring work hour controls for 

individuals performing incidental 

duties, consistent with § 26.205(b)(5), 

from an offsite location in instances 

where those duties might be considered 

to be ‘‘directive’’ in nature. Under the 

proposed amendments to part 26, the 

exclusion of incidental duties while 

calculating work hours would still be 

applicable for licensees and other 

entities licensed under part 53. 

However, for these licensees and other 

entities, beyond instances of incidental 

duties, the direction of operations and 

maintenance activities associated with 

safety-significant SSCs, when performed 

at remote facilities, would be considered 

in an equivalent fashion as direction 

performed at non-remote facilities, for 

the purposes of administering work 

hour controls. 

Proposed § 26.4(b) would include in 

an FFD program individuals who are 

granted unescorted access to the 

protected area of a facility licensed 

under part 53 and do not perform or 

direct the performance of the duties 

described in § 26.4(a). This requirement 

would contribute to the defense-indepth regulatory framework that helps 

provide that individuals who have 

unescorted access are fit for duty, 

trustworthy, and reliable. For example, 

the NRC is proposing amendments to 

part 73 to require a part 53 licensee to 

subject individuals to a series of reviews 

to help determine whether those 

individuals are trustworthy and reliable 

before granting them unescorted access 

to the facility’s protected area. 

The NRC would amend § 26.4(c) to 

include in an FFD program individuals 

who are assigned to physically report to 

the part 53 licensee’s emergency 

response facility (or facilities) or 

participate remotely in emergency 

response activities, and individuals 

without unescorted access to the part 53 

facility who, remotely or otherwise, 

make decisions and/or direct actions 

regarding plant safety or security. Part 

53 commercial nuclear plants may be 

licensed for and rely upon offsite 

facilities to fulfill the role of a Technical 

Support Center or Emergency 

Operations Facility. Therefore, the 

proposed rule would account for such 

offsite facilities or remotely performed 

activities. Further, the use of personnel 

to operate systems and components, 

maintain and surveil SSCs, and respond 

to plant conditions and security events 

may be different than those included in 

the Technical Support Center or 

Emergency Operations Facility team for 

power reactors currently licensed under 

part 50 or part 52. 

For the individuals whose duties for 

the licensees and other entities in 

§ 26.3(c) require the individuals to have 

the types of access or perform the 

activities listed in § 26.4(e)(1) through 

(6) at the location where the commercial 

nuclear plant will be constructed and 

operated, current § 26.4(e) requires them 

to be subject to an FFD program that 

satisfies all the requirements of part 26 

except subparts I and K. The NRC 

would amend § 26.4(e) to except subpart 

M as well as subparts I and K. The NRC 

would also amend § 26.4(e) to include 

in an FFD program the individuals 

whose duties for the licensees and other 

entities in § 26.3(f) require the 

individuals to have the types of access 

or perform the activities listed in 

§ 26.4(e)(1) through (6) or perform 

construction activities as defined in 

§ 26.5. 

Section 26.4(e)(4) would be revised to 

include in an FFD program individuals 

who witness or determine inspections, 

tests, and analyses certifications 

required under part 53 because current 

§ 26.4(e)(4) includes the individuals 

who perform the same duties under part 

52. 

The proposed rule would amend 

§ 26.4(f) to require individuals who 

construct or direct the construction of 

safety- or security-related SSCs at 

facilities licensed under part 53 to be 

subject to an FFD program under 

subpart M of part 26 or an FFD program 

that demonstrates compliance with all 

of the requirements of part 26 except for 

subparts I, K, and M of part 26. 

Section 26.4(g) is the applicability 

paragraph for FFD program personnel 

(e.g., the FFD manager, MRO, and 

technicians) and persons who perform 

AA determinations (e.g., the licensee- or 

other entity-designated Reviewing 

Official). This section would be 

amended to address part 53 licensed 

facilities. Specifically, a part 53 licensee 

or other entity would use FFD program 

personnel to implement its FFD 

program as well as other assigned 

individuals who are not involved in the 

day-to-day operations of the program to 

implement specific elements of its FFD 

program, such as the collection of a 


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specimen for drug or alcohol testing. 

These individuals would be held 

accountable for program 

implementation, including consistent 

implementation of protections afforded 

to all individuals subject to the FFD 

program. 

Section 26.4(h) would be amended to 

include subpart M of part 26. 

The NRC proposes to include several 

new definitions in § 26.5, ‘‘Definitions,’’ 

and amend some existing definitions. 

The NRC is proposing to add a 

definition for ‘‘Biological marker.’’ The 

proposed definition would be consistent 

with ‘‘Biomarker’’ defined by the HHS 

in its Mandatory Guidelines for Federal 

Workplace Drug Testing (HHS 

Guidelines) using oral fluid as the 

biological specimen to be tested (84 FR 

57554; October 25, 2019). However, the 

proposed definition for § 26.5 would 

add that the endogenous substance used 

to validate that the biological specimen 

‘‘was produced by the donor’’ because 

subpart M of part 26 proposes to have 

the MRO evaluate any discrepant 

biological marker identified in a 

biological specimen collected from a 

donor. 

The NRC is proposing a definition for 

the word ‘‘Change’’ as used in the 

proposed § 26.603(e), ‘‘FFD program 

change control,’’ process. The proposed 

definition would be consistent with the 

definition of ‘‘Change’’ for a part 50 or 

52 licensee’s emergency plans in 

§ 50.54(q)(1)(i). 

The NRC proposes to revise the 

definition of ‘‘Constructing or 

construction activities’’ to clarify that 

for licensees or other entities in 

§ 26.3(f), the definition of 

‘‘Construction’’ would be that as 

proposed in § 53.020. 

The definitions of ‘‘Contractor/ 

vendor’’ (C/V) and ‘‘Other entity’’ would 

be revised to make them applicable to 

part 53 licensees. A holder of an ML 

under part 53 could be a C/V under the 

proposed C/V definition. 

The NRC is proposing a definition for 

‘‘Illicit substance’’ because this phrase is 

used in subpart M of part 26 and would 

address substances that cause 

impairment and possible addiction but 

are not an ‘‘illegal drug’’ as defined in 

§ 26.5. This proposal is based on 

operating experience where individuals 

have admitted to using common 

household, non-drug substances to 

achieve a high or satisfy an addiction. 

These common household items 

include, but are not limited to nitrous 

oxide, butane, propane, glue, paint 

vapors, lighter fluid, nail polish 

remover, degreasers, permanent 

markers, and methyl alcohol (which is 

found in hand sanitizer and 

mouthwash). 

The definition of ‘‘Questionable 

validity’’ would be revised to make it 

applicable to an FFD program 

implemented under subpart M of part 

26, which would include all biological 

specimens. 

The NRC is proposing a definition for 

‘‘Reduction in FFD program 

effectiveness’’ because this phrase, 

similar to the proposed definition for 

‘‘Change,’’ is used in proposed 

§ 26.603(e). The proposed definition is 

generally consistent with the definition 

of ‘‘Reduction in effectiveness’’ 

provided for emergency plans in 

§ 50.54(q)(1)(iv). 

The proposed rule would make the 

current definition of ‘‘Reviewing 

official’’ applicable to those licenses and 

other entities in § 26.3(f). 

The current part 26 definition of 

‘‘Safety-related structures, systems, and 

components’’ would be amended to use 

the NRC’s proposed definition in 

§ 53.020 for the part 53 licensees and 

other entities described in § 26.3(d) and 

(f). 

The NRC would amend the definition 

of ‘‘Security-related SSCs’’ in § 26.5 to 

make it applicable to a licensee or other 

entity described in § 26.3(d) and (f). 

The NRC proposes a definition for 

‘‘Special Nuclear Material’’ that would 

refer to the definition in § 70.4, 

‘‘Definitions,’’ of part 70 to ensure 

consistency. 

The NRC is proposing a revision of 

the definition of ‘‘Unit outage’’ to 

account for the potential use of 

commercial nuclear plants for purposes 

other than electricity generation. 

Section 26.21, an applicability 

statement for part 26 FFD programs, 

would be amended to include licensees 

and other entities described in § 26.3(f) 

that choose to implement an FFD 

program that implements all part 26 

requirements, except those in subparts 

K and M of part 26. 

Section 26.51, ‘‘Applicability,’’ would 

be amended to apply to licensees and 

other entities described § 26.3(f) that 

elect not to implement the requirements 

in subpart M of part 26 for the categories 

of individuals in § 26.4 and those 

licensees and other entities that elect to 

implement the requirements in § 26.605. 

Section 26.53(e), (e)(1) and (3), and (g) 

through (i), which are general 

provisions for granting and maintaining 

authorization, would be amended to 

apply to licensees and other entities 

described § 26.3(f). 

Section 26.63(d), a suitable inquiry 

requirement, would be amended to 

apply to licensees and other entities 

described § 26.3(f). 

Section 26.73, the applicability 

statement for subpart D of part 26, 

would be amended to apply to licensees 

and other entities described § 26.3(f) 

that elect not to implement the 

requirements in subpart M of part 26 for 

the categories of individuals in § 26.4 

and those licensees and other entities 

that elect to implement the 

requirements in § 26.605(b). 

Section 26.81, the purpose and 

applicability statement for subpart E of 

part 26, would be amended to apply to 

licensees and other entities described in 

§ 26.3(f) that elect not to implement the 

requirements in subpart M of part 26 for 

the categories of individuals in § 26.4 

and those licensees and other entities 

that implement proposed § 26.605(a) or 

(b). The subpart E requirements to be 

implemented are listed in proposed 

§ 26.607(c)(2)(i) and (c)(2)(ii) and (c)(3). 

Section 26.201, the applicability 

statement for subpart I of part 26 would 

be amended to apply to licensees and 

other entities described in § 26.3(f). 

Also, the applicability statement would 

be divided into two paragraphs for 

clarity. 

The NRC proposes to add § 26.202, 

‘‘General provisions for facilities 

licensed under part 53,’’ for licensees or 

other entities described in proposed 

§ 26.3(f) that elect to implement the 

requirements in subpart I of part 26 in 

accordance with § 26.604 and § 26.605. 

Section 26.202 would establish 

requirements equivalent to those in 

current § 26.203, ‘‘General provisions,’’ 

which is applicable to part 50 and 52 

licensees. The NRC would add the 

separate § 26.202 because § 26.203 refers 

to various requirements under subpart B 

of part 26, which would not be 

applicable to facilities licensed under 

part 53 that implement subpart M of 

part 26. 

Additionally, § 26.202(c), ‘‘Training 

and assessments,’’ unlike § 26.203(c), 

‘‘Training and examinations,’’ would 

not include a comprehensive 

examination requirement because 

trainee assessment is conducted as part 

of a SAT that would be required as 

proposed under the FFD program 

training requirements in § 26.608. 

Proposed changes in §§ 26.205, 

26.207, and 26.211 would add 

references to new requirements in 

subparts I and M of part 26 that would 

be applicable specifically to licensees 

and other entities in § 26.3(f). The NRC 

would not change the specific 

provisions for work hour requirements 

in current § 26.205(d). However, as 

addressed in the discussion of proposed 

changes to § 26.4(a), whether a licensee 

or other entity under part 26 would 

need to implement work hour controls 


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for certain individuals or groups would 

be dependent, in part, on 

determinations reached by that 

licensee’s risk-informed evaluation 

process. 

Proposed changes to 

§§ 26.207(a)(1)(ii) and 26.211(b) would 

allow licensees and other entities in 

§ 26.3(f) to perform face-to-face 

assessments to support the approval of 

work hour control waivers and the 

conduct of fatigue assessments, 

respectively, using electronic 

communications. These proposals 

would allow supervisors to conduct 

such assessments from a remote location 

under appropriate circumstances. Such 

remotely conducted assessments would 

need to be supported by someone who 

is present in-person with the individual 

being assessed and who is trained in 

accordance with the requirements of 

either § 26.29 and § 26.203(c) or § 26.608 

and § 26.202(c). The reasoning for these 

proposals and the associated need for 

in-person support to augment electronic 

communications is addressed further in 

the preamble discussion of proposed 

§ 26.619. 

C. Proposed Requirements for Part 26, 

Subpart M 

The proposed rule would add a new 

subpart M to part 26 that would provide 

alternative FFD requirements for part 53 

licensees and other entities. 

Proposed § 26.601 would make 

subpart M of part 26 applicable to part 

53 licensees and other entities, at their 

discretion. If a licensee or other entity 

in § 26.3(f) does not elect to implement 

an FFD program that demonstrates 

compliance with the requirements of 

subpart M, then the individuals 

specified in § 26.4 would be subject to 

an FFD program that demonstrates 

compliance with all part 26 

requirements, except for those 

requirements in subparts K and M. 

Proposed § 26.603(a) would require an 

applicant to provide a description of its 

FFD program and its implementation 

within its application for a license. This 

requirement is equivalent to the existing 

requirements in §§ 26.401(b) and 

52.79(a)(44). The entities that would be 

required to submit these FFD program 

descriptions are certain applicants that 

would comply with the part 53 

application requirements in subpart H. 

In subpart H, § 53.1309(a)(6) would 

require an applicant for a CP to provide 

a description of its FFD program in its 

PSAR. Under §§ 53.1279(b)(4), 

53.1369(x), and 53.1416(a)(24), an 

applicant for an ML, OL, and COL, 

respectively, would be required to 

provide a description of its FFD 

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 § 26.603(a)(1) would require 

a summary description of the analysis 

described in § 26.603(c), if performed. 

The analysis should describe the 

operation of the facility. This would 

include informing the Commission of: 

(1) the principal individuals assigned by 

job title (work category) and a summary 

description of the human actions (e.g., 

monitoring, operating, responding, 

surveillance, oversight, etc.) that they 

perform to maintain the facility in a safe 

operating or shutdown condition; (2) the 

principal individuals by job title and a 

summarized description of the human 

actions to secure and protect the facility 

(without providing sensitive 

information); (3) the estimated total 

population of individuals subject to the 

FFD program and per shift by job 

description; and (4) references to 

supporting documentation. The purpose 

of these descriptions is to enable an 

NRC assessment of the licensee’s or 

other entity’s analysis and the required 

human actions to operate, monitor, 

surveil, maintain, and secure the facility 

within its design and licensing basis so 

that if an operational or security-related 

event were to occur, the facility would 

respond as designed and licensed and 

the calculated radiological dose 

consequences would not exceed the 

consequences described in 

§ 53.860(a)(2). This is important because 

facilities that implement § 26.604 are 

expected to have very small staff sizes 

and may be sited in geographically 

remote locations, both of which could 

challenge effective implementation of 

the FFD program. 

Proposed § 26.603(a)(2) would require 

the applicant to state what FFD program 

it plans to implement. 

Proposed § 26.603(a)(3) would require 

a discussion that informs the NRC of the 

applicability of the applicant’s FFD 

program to individuals who perform 

safety- or security-significant activities. 

This description should summarize any 

key differences between the staff at the 

site and any remote facility and the 

categories of individuals in § 26.4. The 

principal purpose of providing this 

description would be to inform the NRC 

of any substantial differences in the 

applicability of the FFD program to the 

categories of individuals in § 26.4. 

Proposed § 26.603(a)(4) would require 

a description of the drug and alcohol 

testing and fitness determination 

process to be implemented through the 

licensee’s or other entity’s procedures, 

including the collection and testing 

facilities to be used, biological 

specimens to be collected, and sanctions 

to be imposed upon a confirmed FFD 

policy violation. This process includes 

how individuals who test positive for a 

drug or alcohol will be evaluated before 

being afforded unescorted access to the 

protected area to perform or direct those 

duties or responsibilities making them 

subject to the FFD program. The 

principal purpose of describing this 

return-to-duty process is to inform the 

NRC of the behavioral observation 

strategy (for those facilities that 

implement § 26.604) and/or drug 

screening and testing strategy. 

Proposed § 26.603(a)(5) would require 

a summary description of the 

applicant’s planned PMRP. This 

description must provide the 

performance measures and thresholds 

that the applicant intends to use. 

Proposed § 26.603(b) would establish 

when the FFD program must be 

implemented and the longevity of the 

FFD program. This proposal is 

equivalent to the current § 26.3, which 

states, in part, when licensees and other 

entities must begin implementing their 

FFD programs. Unlike the current part 

26 regulations, proposed § 26.603(b) 

would expressly state that an FFD 

program would not be applicable during 

decommissioning of a part 53 facility for 

licensees and other entities specified in 

§ 26.3(f). However, licensees of facilities 

licensed to operate a reactor should be 

aware that the physical protection 

program under § 73.55, ‘‘Requirements 

for physical protection of licensed 

activities in nuclear power reactors 

against radiological sabotage,’’ and 

under proposed § 73.100 include a 

requirement for the implementation of 

an IMP, even during decommissioning. 

Proposed § 26.603(b) would also 

require the holder of an ML to 

implement its FFD program no later 

than the start of activities that assemble 

a reactor. The holder of the ML should 

establish in its procedures when reactor 

assembly commences and what 

constitutes assembly. For example, the 

FFD program would not need to be 

implemented for the receipt, storage, 

inspection, and staging of components 

and systems used to assemble (i.e., build 

or fabricate) the reactor because this is 

not a current requirement for LWR 

facilities licensed under part 50 or 52. 

Furthermore, the NRC currently does 


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not require that an FFD program be 

applied to the assembly or 

manufacturing of components (or basic 

components as defined in § 21.3), or 

systems that were fabricated or 

assembled outside the footprint of a 

commercial power reactor, and this 

regulatory position would also apply to 

a manufacturing facility. 

Proposed § 26.603(c) would require 

the applicant, licensee, or other entity 

seeking to implement an FFD program 

under § 26.604 to perform a site-specific 

analysis to determine whether the 

facility and its operation satisfy the 

criterion in § 53.860(a)(2). If the analysis 

is performed and demonstrates that the 

radiological consequences presented by 

the facility and its operation satisfy the 

criterion, then the licensee or other 

entity could implement the FFD 

program detailed in § 26.604. If the 

analysis does not demonstrate that the 

facility and its operation satisfy the 

criterion, then the licensee or other 

entity must implement the FFD program 

described in either § 26.605 or subparts 

A through I, N, and O of part 26. 

Proposed § 26.603(c) would also 

require licensees and other entities that 

implement proposed § 26.604 to update 

the technical analysis used to justify 

compliance with the criterion in 

§ 53.860(a)(2). This analysis would be 

updated to reflect changes made to the 

staffing, FFD programs, or offsite 

support resources described in the 

analysis to show that the facility and its 

operation continue to satisfy the 

criterion. This is important because 

facility, operation, or staffing changes 

outside FFD program implementation 

(e.g., changes in the facility safety 

analysis, physical protection strategies, 

or the security plan, implementing 

procedures, or contingency response 

strategies) could adversely impact the 

licensee’s or other entity’s documented 

analysis demonstrating that the facility 

and its operation satisfy the criterion if 

event sequences require human action. 

Proposed § 26.603(d) would require 

the establishment of a PMRP. The 

concept of a PMRP is not new. This 

requirement would consolidate for part 

53 the requirements in current §§ 26.41, 

‘‘Audits and corrective actions’’; 26.415, 

‘‘Audits’’; 26.717, ‘‘Fitness-for-duty 

program performance data’’; and 

26.183(c), which describes MRO 

responsibilities. The proposal would 

state that the licensee or other entity 

must monitor the effectiveness of its 

FFD program by comparing performance 

data against performance measures and 

thresholds. The development of 

quantitative thresholds would be new, 

but this is born from licensees and other 

entities with facilities licensed under 

parts 50 or 52 already collecting, 

reviewing, and reporting FFD 

performance data. Additionally, the 

benefit of quantitatively measuring FFD 

program performance against 

established thresholds benefits a 

licensee’s and other entity’s 

determination of whether they are 

maintaining FFD program performance 

in a manner that demonstrates 

compliance with the performance 

objectives in § 26.23. 

The NRC is proposing the PMRP 

because the subpart M of part 26 

requirements would enable a high 

degree of flexibility in FFD program 

implementation (e.g., drug testing). A 

licensee or other entity would not only 

have options in the type of FFD program 

they may implement under part 26, but 

they would have options in the types of 

biological specimens they may test for 

drugs, where to collect the biological 

specimens (e.g., at the NRC-licensed 

facility or offsite at a local hospital or 

clinic), and the use of collection and 

assessment devices to screen 

individuals for drugs and alcohol. These 

FFD program flexibilities could cause 

FFD programs under subpart M of part 

26 to become very site-specific, 

necessitating performance measures to 

enable the licensee or other entity to 

maintain the effectiveness of its FFD 

program. 

Fitness-for-duty program effectiveness 

would be determined by comparing 

actual performance against the 

performance measures and thresholds. 

The result of that comparison would 

inform licensee or other entity decisions 

whether to change FFD program 

elements to address a performance 

deficiency. Also, the thresholds would 

have sufficient margin, based on 

operating experience, before conditions 

adverse to safety and security may occur 

should an individual be identified as 

impaired or not trustworthy and 

reliable. The potential of a humanrelated failure causing a condition 

adverse to safety and security is 

dependent on the duties and 

responsibilities of the individual and 

the defense-in-depth designed to 

prevent or mitigate an adverse 

consequence. The PMRP would account 

for this by requiring the review of FFD 

performance data, in part, by work 

category, C/V, and individuals 

employed by the licensee who are not 

a C/V as defined in § 26.5 (i.e., a 

licensee employee). 

Proposed § 26.603(d)(1) would require 

the licensee or other entity to document 

and maintain its PMRP. Proposed 

§ 26.603(d)(1)(i) would require that the 

performance measures be identified and 

designed to monitor FFD program 

performance. Proposed 

§ 26.603(d)(1)(i)(A) would require the 

FFD program of a licensee or other 

entity subject to the requirements of 

§ 26.604 to include monitoring of the 

BOP. The purpose of this monitoring is 

to help ensure that individuals subject 

to the FFD program are observing the 

behaviors of others, are being observed 

themselves, and are reporting FFD 

concerns to licensee- or other entitydesignated individuals. The other 

performance measures would include 

occurrence of FFD policy violations 

evaluated by licensee employee, C/V, 

and labor category, and occurrence of 

individuals with potentially 

disqualifying information or who 

possessed an FFD prohibited item. 

Proposed § 26.603(d)(1)(i)(B) would 

require the FFD program of a licensee or 

other entity that is either subject to the 

requirements of § 26.604 and has 

implemented a drug testing program at 

its discretion, or is subject to the 

requirements of § 26.605, to include the 

performance measures identified in 

§ 26.603(d)(1)(i)(A) and those necessary 

to monitor the effectiveness of the drug 

and alcohol testing program. The drug 

and alcohol measures would include the 

monitoring of FFD performance data for 

pre-access and random testing and 

subversion attempts by the categories of 

licensee employee, C/V, and labor 

category. 

Proposed § 26.603(d)(1)(ii) would 

require the licensee or other entity to 

establish thresholds for each 

performance measure. Initial thresholds 

must be based on FFD performance data 

from comparable facilities subject to 

part 26, the licensee’s or other entity’s 

fleet-level program performance if 

applicable, and industry FFD 

performance data. This provision 

introduces the requirement to ‘‘maintain 

FFD program effectiveness.’’ This 

terminology describes a performancebased regulatory strategy in which the 

licensee or other entity must initially 

establish a level of performance that is 

representative of other facilities in the 

licensee’s fleet of facilities subject to 

part 26, if applicable, and the FFD 

performance of comparable facilities 

subject to part 26. 

Proposed § 26.603(d)(1)(iii) would 

require that the licensee or other entity 

evaluate FFD data as it is received to 

determine whether a threshold has been 

exceeded. Historical FFD performance 

data for the current LWR fleet indicates 

that, for particular work categories and 

employment types, few FFD policy 

violations occur per year. Therefore, for 

work categories that may be significant 

to worker safety (e.g., radiation 

protection technicians), physical 


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protection (i.e., security personnel), or 

safety (i.e., NRC-licensed operators and 

individuals who perform or direct the 

performance of activities that a riskinformed evaluation process has shown 

to be significant to public health and 

safety), a single FFD policy violation 

could be a significant occurrence and 

warrant corrective actions. Based on 

licensee-submitted FFD-related reports 

under §§ 26.417, 26.419, 26.717, and 

26.719, licensees and other entities with 

facilities licensed under parts 50 or 52 

implement some form of corrective 

action that is typically scaled to the 

significance of the violation. These 

corrective actions have included 

counseling, follow-up drug and/or 

alcohol testing, remedial training, 

generic announcements to the 

workforce, and reviews of recently 

performed or directed work by the 

individual suspected of being impaired. 

Proposed § 26.603(d)(1)(iii) would 

require that the PMRP include a year-toyear comparison of FFD performance 

data to help provide assurance that an 

adverse trend in FFD program 

performance would be identified if 

occurring. This proposed requirement 

was developed from the annual FFD 

performance data reporting 

requirements in §§ 26.417(b)(2) and 

26.717. In particular, the proposed yearto-year comparison of FFD performance 

data is equivalent to § 26.717(c), which 

requires, in part, licensees and other 

entities to analyze their performance 

data at least annually and take 

appropriate actions to correct any 

identified program weaknesses. 

Proposed § 26.603(d)(1)(iv) would 

require the licensee or other entity to 

perform and document quantitative and 

qualitative reviews. These reviews 

would be performed in three program 

areas: protections afforded to 

individuals subject to the FFD program, 

laboratory test results and MRO 

performance, and change control. The 

purpose of these reviews would be to 

specifically target performance within 

the three program areas to assess 

whether the outcomes resulting from the 

implementation of procedure 

requirements are contributing to FFD 

program effectiveness. The proposed 

reviews would not require the 

establishment of measures and 

thresholds because the reviews are 

expected to result in qualitative findings 

regarding program effectiveness. 

Qualitative findings and observations 

could still result in the consideration of 

corrective actions in the targeted 

program areas. 

Proposed § 26.603(d)(1)(iv)(A) would 

require the licensee or other entity to 

monitor whether its FFD program is 

affording appropriate protections to 

individuals subject to the FFD program. 

The review of these protections would 

include, in part, assessing the licensee’s 

or other entity’s protection of the 

following: privacy during the specimen 

collection process; specimen integrity, 

custody, and control; information 

gathered from FFD program 

implementation; and due process during 

appeals of FFD policy violations. 

Proposed § 26.603(d)(1)(iv)(B) would 

require, in part, a review of laboratory 

test results and MRO performance. 

Effective performance by the laboratory 

(e.g., obtaining and communicating 

accurate test results) and MRO (e.g., 

correct evaluation of the laboratory test 

results based on § 26.185 or HHS 

Guidelines) would result in three 

significant outcomes: (1) protection of 

the donor from an inaccurate FFD 

policy violation determination; (2) 

protection of the donor, other 

individuals, and the facility from 

potential harm should the donor be 

impaired or not trustworthy and 

reliable; and (3) a performance-based 

assessment of both the laboratory and 

MRO. This last outcome could facilitate 

actions to improve laboratory 

performance, MRO training under 

§ 26.607(m), or both. Proposed 

§ 26.603(d)(1)(iv)(B) would also require 

a comparative analysis between the 

POCTA screening result(s) and the 

corresponding specimen test results 

obtained from the HHS-certified 

laboratory if the POCTA indicated a 

positive, adulterated, substituted, or 

invalid screening result or discrepant 

biological marker, to assess the 

effectiveness of the POCTA and to 

inform MRO decisions under § 26.185 or 

§ 26.607(m)(6). The results of this 

biennial review could also inform the 

conduct of laboratory audits. 

Proposed § 26.603(d)(1)(iv)(C) would 

require that the change control 

requirement in proposed § 26.603(e) be 

included in the biennial program review 

to help ensure that changes 

implemented over the life of the facility 

do not result in a reduction in program 

effectiveness even if a mitigating action 

was implemented for the specific 

change. This requirement was 

developed from §§ 26.137(f) and 

26.713(d). This part of the review would 

require an assessment of all changes 

since the last review and their potential 

aggregated impact on FFD program 

effectiveness. For example, if last year 

the licensee elected to contract with a 

different MRO and this year the licensee 

implemented a new type of POCTA 

device, each of those program changes 

probably would not have resulted in a 

recognizable reduction in FFD program 

effectiveness. But, if the drug testing 

positivity rate (or FFD policy violations) 

for C/Vs decreased markedly during a 

future maintenance outage that required 

many C/Vs, then the reduction could 

indicate, for example, that the POCTA 

device was not as effective as 

determined by a forensic toxicologist 

review under §§ 26.603(e) and 26.607(h) 

or that the new MRO was improperly 

crediting prescription medication for 

laboratory-confirmed positive test 

results. 

Proposed § 26.603(d)(2) would state 

when the licensee or other entity must 

implement corrective actions. This 

requirement would be equivalent to the 

requirement in current § 26.415(b) and 

was developed from requirements 

contained in §§ 26.41(a) and (f), 

26.127(e), 26.129(b)(1)(i), 26.137(f)(3) 

through (5), 26.155(a)(6), 26.157(e), 

26.159(b)(1)(i), and 26.203(e)(2). 

Corrective actions must be implemented 

to correct root causes, contributing 

causes, or both. There is margin built 

into the FFD performance thresholds 

and qualitative factors (e.g., to account 

for potential changes in drug and 

alcohol testing performance data when 

there is a large influx of C/Vs to perform 

maintenance) that may influence a 

licensee or other entity’s causal 

determination for an occurrence. Thus, 

generalized or qualitative corrective 

actions may be implemented like 

informing management and placing a 

sufficiently descriptive summary of the 

occurrence in a corrective action 

program for future monitoring to assess 

recurrence. 

However, should the occurrence 

challenge safety or security or 

significantly exceed a performance 

threshold even when considering 

qualitative factors and margin, the 

licensee or other entity should 

implement more robust corrective 

actions to resolve the cause. An example 

of a challenge to safety or security 

would be the situation when an NRClicensed operator or maintenance 

professional had operated, surveilled, or 

maintained safety-significant SSCs and 

was determined to have been impaired 

by behavioral observation or potentially 

under the influence of a narcotic as 

determined by an alcohol or drug test or 

screening result. Immediate corrective 

actions could include, but would not be 

limited to, a licensee or other entity 

assessment of the duties and 

responsibilities recently performed by 

the individual. Operating experience 

within the LWR operating reactor 

community demonstrates few FFD 

policy violations per year per site have 

been caused by individuals who 

perform or direct the performance of 


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10The Drug Enforcement Administration 

classifies drugs, substances, and certain chemicals 

used to make drugs into five (5) distinct categories, 

depending upon the drug’s acceptable medical use 

and the drug’s abuse or dependency potential. 

These categories appear as Schedules I through V 

of section 202 of the Controlled Substances Act (21 

U.S.C. 812). Schedule I drugs have a high potential 

for abuse, have no currently accepted medical uses 

in treatment in the United States, and lack accepted 

safety for use under medical supervision. At the 

other end of the classification scheme, Schedule V 

drugs have the least potential for abuse among the 

five categories of drugs, have a currently accepted 

medical use in treatment in the United States, and 

abuse of the drug may lead to limited physical 

dependence or psychological dependence. For more 

information, see https://www.dea.gov/druginformation/drug-scheduling. 

safety or security-significant activities. 

Therefore, any such violations of the 

FFD policy in a particular work category 

in one year could be a significant 

performance deficiency. These 

violations could be even more 

significant at part 53 facilities that have 

a very small workforce subject to part 

26. 

Proposed § 26.603(d)(3) would require 

the licensee or other entity to biennially 

assess and document its FFD 

performance monitoring program; this 

requirement was developed from 

§ 26.41(b). This documented review 

would demonstrate that the 

performance measures and thresholds 

are appropriate based on site- and 

licensee’s fleet-level program 

performance, if applicable, and industry 

performance and adjusted to maintain 

FFD program effectiveness. Also, as a 

result of this effort, the licensee or other 

entity would be in possession of lessons 

learned from fleet-level performance, if 

applicable, and industry performance 

that could contribute to their own 

performance assessment to maintain 

program effectiveness. 

Under proposed § 26.603(d)(3)(i), the 

identified program weaknesses and 

corrective actions resulting from the 

biennial review would be required to be 

summarized in the licensee’s or other 

entity’s annual report to the NRC in 

compliance with either § 26.417(b)(2) or 

§ 26.717, as applicable. This information 

would inform the NRC of FFD program 

weaknesses to facilitate regulatory 

oversight and enable the NRC to 

aggregate industry data for use in a 

licensee or other entity PMRP. 

Proposed § 26.603(d)(3)(ii) would 

establish when the biennial PMRP 

review must be completed and when 

corrective actions from the review must 

be implemented. The NRC selected the 

May 15th date of odd-numbered years to 

help ensure that all FFD programs will 

maintain their previously determined 

performance measures and thresholds or 

reset them based on FFD program 

performance early in the year in which 

the biennial review was conducted. This 

would assist in obtaining quality FFD 

performance data over two annual 

reporting cycles and evaluating whether 

previous corrective actions were 

effective. 

In proposed § 26.603(e), the NRC 

proposes a change control requirement 

for subpart M of part 26 FFD programs. 

Requiring licensees and other entities to 

demonstrate compliance with certain 

requirements before implementing 

changes to their FFD programs would be 

necessary for two primary reasons. First, 

proposed changes to a licensee’s or 

other entity’s FFD program could affect 

the analysis performed by the licensee 

or other entity under proposed 

§ 26.603(c), which helps determine the 

FFD program requirements that must be 

implemented. If this analysis changes, 

then the licensee’s or other entity’s FFD 

program requirements might change. 

Second, the requirements in subpart M 

of part 26 are performance based. 

Therefore, FFD program implementation 

may change periodically in response to 

societal changes in substance abuse or 

from PMRP implementation. Change 

control therefore relies on the licensee 

or other entity maintaining its 

procedures in a manner that details how 

its FFD program is to be implemented 

while incorporating changes, with 

documentation that justifies the changes 

to support the PMRP, audits, and NRC 

inspection. 

Proposed § 26.603(e)(1) would permit 

the licensee or other entity to 

implement changes to its FFD program 

if it performs and retains an analysis 

demonstrating that the change does not 

reduce the effectiveness of the FFD 

program or the change was necessitated 

or justified by a change to part 26, 

laboratory processes, or guidance issued 

by the HHS or NRC. The proposed 

change control requirement would 

enable flexibility in program 

implementation should the NRC or HHS 

change its drug testing procedures (as 

implemented by the licensee or other 

entity through its procedures) in 

response to changes in societal 

substance abuse or drug testing 

technologies. 

The proposed change control 

requirement was developed from the 

change control requirements in 

§ 50.54(p) and (q)—the change control 

requirements for security and 

emergency plans, respectively. 

However, unlike these two 

requirements, the NRC does not review 

and approve a licensee’s or other 

entity’s FFD program or its 

implementing procedures, and the FFD 

program is not licensing-basis 

information as described in § 53.1300. 

Proposed § 26.603(e)(2) would require 

that if a change reduces FFD program 

effectiveness, then the licensee must 

implement a mitigating strategy so the 

FFD program, as revised, will continue 

to demonstrate compliance with the 

performance objectives in § 26.23 and 

not result in a reduction in program 

effectiveness. 

Proposed § 26.603(e)(3) would 

prohibit, with one exception, the use of 

the change control process to reduce the 

minimum panel of drugs to be tested 

and would reference the drugs listed in 

proposed § 26.607(c)(1). Proposed 

§ 26.607(c)(1) would reference current 

§ 26.31(d)(1), which states that, at a 

minimum, licensees and other entities 

shall test for marijuana metabolite, 

cocaine metabolite, opioids (codeine, 

morphine, 6-acetylmorphine, 

hydrocodone, hydromorphone, 

oxycodone, and oxymorphone), 

amphetamines (amphetamine, 

methamphetamine, 

methylenedioxymethamphetamine, and 

methylenedioxyamphetamine), 

phencyclidine, and alcohol. The testing 

of these drugs and drug metabolites, 

except phencyclidine, and alcohol is 

necessary for the FFD program to 

remain effective. Also, there is no 

proposed subpart M of part 26 

requirement stating that this panel of 

drugs and drug metabolites needs to 

consist of only scheduled drugs.10 This 

flexibility would account for the 

situation where an impairing substance 

becomes prevalent in society and a 

licensee or other entity elects to add the 

substance to their panel of substances to 

be tested prior to it being scheduled by 

the Drug Enforcement Administration. 

The exception in proposed 

§ 26.603(e)(3) would be that, should 

HHS elect to remove phencyclidine 

from the panel of drugs and drug 

metabolites to be tested, a licensee or 

other entity could make this change in 

its FFD program without resulting in a 

reduction in FFD program effectiveness. 

This outcome would be justified based 

on the very infrequent occurrence rate 

of FFD policy violations due to 

phencyclidine use since 2010. However, 

if HHS proposes to remove a class of 

drugs from the panel of drugs to be 

tested that is listed in § 26.31(d)(1), 

except for phencyclidine, then a 

licensee or other entity may not make a 

similar change to its panel of drugs to 

be tested, because this change would be 

a reduction in FFD program 

effectiveness even with a mitigative 

strategy implemented. 

Changes in the HHS panel of drugs 

and drug metabolites to be tested may 

also shift from one metabolite to a 


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different metabolite for the same drug 

class (e.g., amphetamines, opioids) to be 

tested. Should HHS issue such a change 

to its panel, this would not be expected 

to result in a reduction in FFD program 

effectiveness because HHS would be 

targeting a more prevalent or effective 

metabolite in its drug testing program. 

This situation could occur as HHS 

gathers more operating experience from 

Federal Government implementation of 

its HHS Guidelines, or data generated by 

drug testing laboratories and federally 

mandated drug testing programs 

required by Federal agencies such as the 

NRC and U.S. Departments of 

Transportation, Energy, and Defense. 

Proposed § 26.603(e)(4) would require 

that change control records be 

maintained for a 5-year record retention 

period based on the current NRC 

practice to conduct triennial inspections 

of licensees’ and other entities’ FFD 

programs. This would afford the NRC an 

opportunity to review the licensee’s or 

other entity’s determination that FFD 

program changes have not reduced the 

effectiveness of their FFD program. 

Licensees and other entities would also 

be required to summarize each change 

made under proposed § 26.603(e) in 

their annual FFD performance reports 

required by § 26.617(b)(2) or § 26.717, as 

applicable. 

Proposed § 26.604 would establish the 

minimum set of FFD program 

requirements for licensees and other 

entities who have a documented 

analysis that demonstrates that the 

facility and its operation satisfy the 

criterion in § 53.860(a)(2). For these 

licensees, compliance with the 

performance objectives in § 26.23 would 

be ensured through the BOP; defense-indepth measures proposed in subpart M 

of part 26 like the PMRP, change 

control, and audits; and other 

requirements, such as those for AA, 

physical protection, and licensed 

operators. The adequacy of these 

measures in satisfying the performance 

objectives is supported by operating 

experience, which demonstrates margin 

between an FFD-related occurrence and 

a condition adverse to safety or security, 

as illustrated by for-cause, post-event, 

and random testing data. A facility that 

satisfies the criterion in proposed 

§ 53.860(a)(2) would present a smaller 

potential radiological consequence than 

a facility that does not satisfy the 

criterion, so the requirements in 

proposed § 26.604 are scaled to the 

lower risk presented consistent with the 

Commission’s Advanced Reactor Policy 

Statement. 

The disadvantages of implementing 

the FFD program described in proposed 

§ 26.604 would be few. Since drug and 

alcohol testing would not be required, 

behavioral observation would be the 

keystone requirement in this 

performance-based framework to 

provide that individuals are fit for duty, 

trustworthy, and reliable, and can safely 

and competently perform the duties and 

responsibilities making them subject to 

the FFD program. If not, the individuals 

would be assessed in accordance with 

the licensee’s or other entity’s 

procedures similar in manner to that 

required by subpart K of part 26, and the 

proposed PMRP would require 

corrective actions should a threshold be 

exceeded. 

If a licensee or other entity elects not 

to perform the analysis in proposed 

§ 26.603(c) to determine whether it 

satisfies the criterion in proposed 

§ 53.860(a)(2); performs the analysis and 

finds that the facility and its operation 

does not satisfy the criterion in 

proposed § 26.603(c); or is a holder of an 

ML, the licensee or other entity could 

not implement the FFD program 

described in § 26.604. Instead, the 

licensee or other entity would 

implement either the program described 

in proposed § 26.605 or an FFD program 

that demonstrates compliance with all 

the requirements in current subparts A 

through I, N, and O of part 26. 

Proposed § 26.605 would establish 

requirements in a graded manner 

similar to the regulatory framework 

established by the requirements in 

subparts A through I, N, O, and K of part 

26. This existing graded approach 

consists of an FFD program for 

construction of a commercial nuclear 

plant and a more robust program that 

must be implemented before reactor 

operation. The former is the FFD 

program in proposed § 26.605(a), and 

the latter is proposed § 26.605(b). Like 

that for an FFD program under § 26.604, 

the FFD program under § 26.605 would 

include FFD program elements similar 

to those in subpart B of part 26, but the 

proposed requirements are less 

prescriptive, enabling more flexibility in 

program implementation like that 

offered in subpart K of part 26. For 

example, the requirements in subpart B 

of part 26 are explicit requirements for, 

in part, the collection and analysis of 

urine specimens. Subpart B of part 26 

does not enable the use of oral fluid for 

drug testing or screening, except under 

very limited situations as described in 

subpart E of part 26, or the use of hair 

specimens, unlike proposed § 26.605. 

Proposed § 26.605 would require drug 

and alcohol testing based on either the 

requirements in part 26 or the HHS 

Guidelines. The principal benefit of the 

proposed § 26.605 FFD program is that 

it would provide a regulatory framework 

that is consistent with the radiological 

consequences for a facility that does not 

satisfy the criterion in proposed 

§ 53.860(a)(2) while affording 

flexibilities in the conduct of drug and 

alcohol testing. 

Proposed § 26.605(a) would apply to 

licensees and other entities who 

perform the § 26.603(c) analysis and 

satisfy the criterion in § 53.860(a)(2) but 

decide not to implement the FFD 

program described in proposed § 26.604, 

licensees and other entities who do not 

perform the § 26.603(c) analysis, and 

licensees and other entities who 

perform the analysis but their analysis 

does not demonstrate that their facility 

and its operation satisfy the criterion in 

§ 53.860(a)(2). These entities must 

establish, implement, and maintain an 

FFD program under § 26.605(a) either 

during construction activities as defined 

in § 26.5, or during activities performed 

under an ML that allows the assembly, 

testing, or both, of a manufactured 

reactor. This FFD program implements 

all the FFD program requirements in 

§ 26.604 plus drug and alcohol testing. 

The timing element of the proposed 

applicability statement of § 26.605(a) is 

equivalent to that for an LWR licensee 

or other entity who is performing those 

same activities at a facility licensed 

under part 50 or 52 and helps provide 

assurance that those individuals who 

assemble, test, or perform construction 

activities as defined in § 26.5 or direct 

these activities are fit for duty and 

trustworthy and reliable. This is 

important because assembly and testing 

a manufactured reactor and the 

construction and testing of SSCs 

required for facility operation require, in 

part, adherence to procedures, possible 

implementation of unique and precise 

assembly techniques, and quality 

assurance and controls. Additionally, 

SSCs within a manufactured reactor 

may not be accessible, testable, or 

available for quality assurance and 

verification after the reactor is 

assembled. This requirement is also 

proposed to address solo-assembly 

activities that may cause latent failures 

and passive SSCs located internal to a 

reactor (for example, a fusible link 

designed to melt at a particular 

temperature to trigger an actuation 

mechanism) that are relied upon for safe 

operation but cannot be inspected or 

tested for proper installation, 

configuration, or operation after 

installation. A § 26.605(a) FFD program 

for these types of activities is equivalent 

to the FFD program applicable to the 

assembly of the reactor vessel internals 

and testing of the SSCs internal to the 

reactor at an LWR licensed under part 

50 or 52. 


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Proposed § 26.605(b) would apply to 

the same licensees and other entities as 

in proposed § 26.605(a) but before the 

loading of fuel onsite into a reactor 

vessel; before receiving a manufactured 

reactor; or before individuals subject to 

part 26 operate, test, perform 

maintenance of, or direct the 

maintenance or surveillance of securityrelated equipment or equipment that a 

risk-informed evaluation process has 

shown to be significant to public health 

and safety. These entities must 

establish, implement, and maintain an 

FFD program that implements all the 

requirements in § 26.605(a), except 

proposed §§ 26.610, ‘‘Sanctions’’; 

26.617, ‘‘Recordkeeping and reporting’’; 

and 26.619, ‘‘Suitability and fitness 

determinations’’; plus additional 

requirements due to the increased 

radiological consequences presented by 

a part 53 commercial nuclear plant as 

the licensee readies it for operation. 

These additional requirements include 

those in subparts C, D, H, and N of part 

26, some of which would replace 

§§ 26.610, 26.617, and 26.619. 

Proposed § 26.605(b) would also 

enable the licensee or other entity to 

better integrate its facility into the LWR 

fleet and Category I fuel cycle facilities 

because subparts C, D, and H of part 26 

would be required. These subparts 

would be required, in part, because it is 

expected that: (1) individuals will be 

able to work at any part 50, 52, or 53 

commercial nuclear plant and will 

possess a nuclear safety culture and 

desirable qualifications, skills, 

expertise, or services; and (2) licensees 

and other entities of facilities licensed 

under parts 50, 52, and 70 may venture 

to construct or operate a facility 

licensed under part 53. Therefore, the 

implementation of these subparts would 

help ensure that all individuals subject 

to part 26, except those individuals 

subject to an FFD program under 

§ 26.604, § 26.605(a), or subpart K of 

part 26, would be subject to FFD 

programs that provide reasonable 

assurance that the individuals are fit for 

duty, trustworthy, and reliable. 

Proposed § 26.606, ‘‘Written policy 

and procedures,’’ would require 

licensees and other entities to 

implement and maintain an FFD policy 

and procedures for their FFD programs. 

This section would establish 

requirements equivalent to those in 

current § 26.403, ‘‘Written policy and 

procedures,’’ of subpart K. However, a 

principal difference is that proposed 

§ 26.606 is written to enable the use of 

urine, oral fluid, and hair for drug 

testing and screening. 

Proposed § 26.606(a)(1) would require 

each licensee and other entity to 

provide a written FFD policy statement 

to individuals subject to the FFD 

program before the individuals are 

subjected to behavioral observation and 

any FFD program drug and alcohol test. 

This would be a protection measure 

afforded to individuals subject to the 

FFD program to help ensure that they 

know what is expected of them before 

being subject to the FFD program and 

potential consequences should they 

violate the FFD policy or procedures. 

This requirement would also contribute 

to safety and security because 

understanding FFD program 

responsibilities may enhance an 

individual’s safety culture or the 

individual may self-select out of the 

licensee’s or other entity’s hiring 

process. 

Proposed § 26.606(a)(2) would require 

that the FFD policy statement describe 

the performance objectives in § 26.23, 

which are the same FFD program 

performance objectives required for 

facilities licensed under parts 50, 52, or 

70. Having a standard performance 

outcome based on a licensee or other 

entity satisfying the § 26.23 performance 

objectives would enhance consistency 

in FFD program implementation across 

all entities subject to part 26. It would 

also generate confidence that 

individuals subject to part 26 will safely 

and competently perform their duties 

and responsibilities and use NRClicensed materials in a manner that will 

protect the public health and safety and 

common defense and security. 

Proposed § 26.606(a)(3) would require 

that the FFD policy statement describe 

the minimum days off requirements in 

§ 26.205(d)(3) or maximum average 

work hours requirements in 

§ 26.205(d)(7). 

Proposed § 26.606(a)(4) would require 

the FFD policy statement be written in 

sufficient detail to provide affected 

individuals with information on what is 

expected of them and what 

consequences may result from a lack of 

adherence to the policy, including those 

elements described in § 26.603(b), part 

26-required sanctions, and required 

medical/clinical treatment and followup testing for FFD policy violations. 

This requirement is equivalent to 

§ 26.403(a) of subpart K but includes an 

additional description of what the 

policy statement must include. For 

example, the policy would describe the 

NRC-required sanctions to help deter 

substance abuse and required medical/ 

clinical treatment and follow-up testing 

for FFD policy violations. This 

provision would provide a protection 

measure by helping the individual get 

the assistance they need and help 

ensure that the individual refrains from 

substance abuse. 

Proposed § 26.606(a)(5) would require 

that the FFD policy statement describes 

the individual’s responsibilities to 

report for work in a physiological and 

psychological condition that enables the 

safe and competent performance of 

assigned duties and responsibilities and 

inform a licensee- or other entitydesignated representative when the 

individual determines that this cannot 

be accomplished. 

Proposed § 26.606(b) would require 

licensees and other entities 

implementing a FFD program in 

accordance with subpart M of part 26 to 

establish, implement, and maintain 

written procedures for their FFD 

programs. This requirement would be 

equivalent to that in § 26.403(b) of 

subpart K. 

Proposed § 26.606(b)(1) would 

establish requirements for a subpart M 

of part 26 FFD program in which the 

licensee or other entity implements a 

drug and alcohol testing program. This 

provision would be equivalent to the 

requirements in current § 26.403(b)(1) of 

subpart K, but § 26.606(b)(1)(i) through 

(iv) proposes additional clarity and 

specificity that licensees and other 

entities must detail in their procedures 

to address new testing methods in 

subpart M of part 26 that are not 

permitted under the current part 26 

framework. Clarity and specificity in 

procedural instructions would support 

consistent program implementation, 

which protects all individuals subject to 

the program. 

Proposed § 26.606(b)(1)(iv) would 

require that if the licensee or other 

entity elects to use the HHS Guidelines 

for the conduct of drug testing, the FFD 

program procedures must include the 

name of the specific HHS Guideline and 

revision being implemented by the 

licensee or other entity and a 

description of the specific sections in 

the guideline that are being 

implemented, including specimen 

collections, drug testing, laboratory 

procedures, and evaluation of test 

results. This requirement would help 

ensure the following: the validity and 

accuracy of drug testing because the 

specimens would be subject to 

laboratory testing that has been certified 

by the HHS; protection of worker rights 

equivalent to the privacy, information, 

and due process protections afforded to 

Federal workers under the HHS 

Guidelines because the HHS Guidelines 

are used in the Federally mandated drug 

testing programs; consistency in 

program implementation because all 

individuals subject to the FFD program 

would be subject to the same collection, 


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11By ‘‘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. 

testing, and evaluation processes; and 

FFD program effectiveness because the 

effectiveness of the HHS Guidelines 

have been verified by HHS’s National 

Laboratory Certification Program 

(NLCP). Detailed procedures would 

enhance MRO and FFD program 

personnel reviews of individual test 

results because instructions would be 

provided for, in part, the evaluation of 

specific test results (e.g., positive, 

negative, biological markers), the 

conduct of additional testing for invalid 

or dilute specimens, and the assessment 

of subversion attempts (e.g., adulterated 

or substituted). This would benefit FFD 

program effectiveness and help prevent 

misunderstanding of program 

requirements and processes. 

Proposed § 26.606(b)(2) would require 

licensees and other entities to include in 

their written procedures the immediate 

and follow-up actions that would be 

taken, and the procedures that would be 

used, in certain situations specified in 

proposed § 26.606(b)(2)(i) through (vi). 

Proposed § 26.606(b)(2) would be 

equivalent to the requirements in 

current § 26.403(b)(2), which provides 

the same requirement under an FFD 

program for construction for part 50 or 

52 licensees and other entities. This 

would help ensure the effectiveness of 

the FFD program and its consistent 

implementation, because part 53 

licensed facilities would be 

implementing procedures to address the 

same requirements and with individuals 

who would understand what is 

expected of them no matter what part 53 

facility they were assigned. 

The situation specified in proposed 

§ 26.606(b)(2)(i) would arise when 

individuals subject to the FFD program 

have been involved in the use, sale, or 

possession of illegal substances, illegal 

drugs, or illicit substances. This 

provision would be equivalent to 

current § 26.403(b)(2)(i), except that the 

phrase ‘‘illegal drugs’’ would be 

replaced with ‘‘illegal substances, illegal 

drugs, or illicit substances.’’ Illegal 

substances would include legal 

substances used in a manner 

inconsistent with Federal or State law. 

The situation specified in proposed 

§ 26.606(b)(2)(ii) would arise when 

individuals who are subject to the FFD 

program are impaired by any substance 

or the consumption of alcohol as 

determined by behavioral observation or 

a test that measures blood alcohol 

concentration, as defined in § 26.5. 

Except for a few differences, this 

provision would be equivalent to 

current § 26.403(b)(2)(ii) of subpart K. 

The NRC would not include the phrases 

‘‘to excess’’ and ‘‘accurately’’ in 

proposed § 26.606(b)(2)(ii). Subpart M of 

part 26 is a performance-based 

framework that focuses on impaired 

human performance, and for alcohol, 

impairment is determined by behavioral 

observation or by blood alcohol 

concentrations exceeding the limits in 

§ 26.103, ‘‘Determining a confirmed 

positive test result for alcohol,’’ using an 

evidentiary breath testing (EBT) device 

for alcohol (not whether an individual 

drank ‘‘to excess’’). If impairment is 

determined by an individual’s behavior, 

it must be based on physiological 

indications of alcohol impairment. 

These indications are well established 

in medical, clinical, and law 

enforcement organizations, and could be 

used by the licensee or other entity 

through its procedures and training.11

The NRC would include the phrase 

‘‘illegal substances, illegal drugs, and 

illicit substances’’ in proposed 

§ 26.606(b)(2)(ii) based on operating 

experience and the terminology in 

current § 26.23(b). There are far more 

substances that may cause impairment 

than just drugs, drug metabolites, and 

alcohol. The phrase ‘‘before or while 

constructing or directing construction of 

safety- or security-related SSCs’’ in 

current § 26.403(b)(2)(ii) would not be 

included in proposed § 26.606(b)(2)(ii) 

because proposed § 26.606 would apply 

during construction, operation, and 

decommissioning, if applicable. The 

NRC would include the term 

‘‘behavioral observation’’ in proposed 

§ 26.606(b)(2)(ii) because impairment 

can be visibly or audibly observed in an 

individual, and individuals subject to 

subpart M of part 26 would be trained 

in behavioral observation under 

proposed § 26.608. 

The situation specified in proposed 

§ 26.606(b)(2)(iii) would arise when 

individuals who are subject to an FFD 

program that includes drug and alcohol 

testing attempt to subvert the testing 

process by adulterating or diluting 

specimens (in vivo or in vitro), 

substituting specimens, or by any other 

means. Except for one difference, this 

provision would be equivalent to 

current § 26.403(b)(2)(iii). The NRC 

would include the phrase ‘‘if drug and 

alcohol testing is conducted’’ to address 

the licensee or other entity who 

implements § 26.604, which does not 

require drug and alcohol testing. The 

purpose underlying this requirement 

has increased in significance since 

issuance of the 2008 part 26 final rule 

because subversion attempts have 

accounted for about one-third of all FFD 

policy violations every year since 2016. 

The situation specified in proposed 

§ 26.606(b)(2)(iv) would arise when 

individuals, who are subject to an FFD 

program that includes drug and alcohol 

testing, refuse to provide a specimen for 

analysis or refuse to follow instructions 

provided by FFD program personnel. 

Except for two differences, this 

provision would be equivalent to 

current § 26.403(b)(2)(iv). As with 

proposed § 26.606(b)(2)(iii), the NRC 

would include the phrase, ‘‘if drug or 

alcohol testing is conducted,’’ to 

account for an FFD program 

implemented under § 26.604. The NRC 

would include the phrase ‘‘or follow the 

instructions provided by FFD program 

personnel’’ based on an existing 

requirement in § 26.89(c) that the 

collector must inform the donor that if 

the donor refuses to cooperate in the 

specimen collection process, then such 

refusal will be considered a refusal to 

test and sanctions for subverting the 

testing process will be imposed. 

The situation specified in proposed 

§ 26.606(b)(2)(v) would arise when 

individuals who are subject to an FFD 

program had legal action taken relating 

to drug or alcohol use. This requirement 

would be equivalent to current 

§ 26.403(b)(2)(v). 

The situation specified in proposed 

§ 26.606(b)(2)(vi) would be when 

individuals subject to an FFD program 

demonstrated character or actions 

indicating that the individual cannot be 

trusted or relied upon to perform those 

duties and responsibilities or maintain 

access to NRC-licensed facilities, SNM, 

or sensitive information. This includes 

character traits beyond those attributed 

to drug or alcohol use. This proposal 

would help ensure that the licensee or 

other entity will implement an FFD 

program designed to demonstrate 

compliance with the § 26.23(c) 

performance objective that FFD 

programs must provide ‘‘reasonable 

measures for the early detection of 

individuals who are not fit to perform 

the duties that require them to be 

subject to the FFD program.’’ An 

individual who is not trustworthy and 

reliable is not fit to perform or direct the 

performance of those duties and 

responsibilities or be afforded those 

types of access that make the individual 

subject to an FFD program. 

This proposed requirement also 

would help to align the subpart M of 

part 26 BOP with the BOP implemented 

under § 73.56(f) and proposed § 73.120 

and the purpose of the IMP as described 

in § 73.55(b)(9) and proposed 


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12The IMP must monitor the initial and 

continuing trustworthiness and reliability of 

individuals granted or retaining unescorted AA to 

a protected or vital area and implement defense-indepth methodologies to minimize the potential for 

an insider to adversely affect, either directly or 

indirectly, the licensee’s capability to protect 

against radiological sabotage. 

§ 73.100(b)(9).12 The demonstrated 

character and actions of an individual 

can indicate whether the individual can 

be trusted and relied upon to safely and 

competently perform assigned duties 

and responsibilities or be afforded those 

types of access making the individual 

subject to the FFD program. This holds 

true for any demonstrated adverse 

character indication or action on- or 

offsite. 

The phrase ‘‘character or actions’’ 

would be used in proposed 

§ 26.606(b)(2)(vi) to focus on observed 

examples that indicate an individual 

subject to subpart M of part 26 may not 

be fit for duty or trustworthy and 

reliable. Character traits include but are 

not limited to personality, temperament, 

honesty, carelessness, apathy, 

psychosis, and commitment to safety 

culture. Assessment of an individual’s 

character should consider the potential 

for changes in these traits when 

compared to a previous baseline. 

Actions would include a physical or 

verbal demonstration of a character trait 

that could call into question an 

individual’s fitness, trustworthiness, or 

reliability. For example, the individual 

does something physically, verbally, or 

in writing (e.g., falsifying records, 

driving while impaired, or harming or 

threatening to harm oneself, others, or 

property) that compels another 

individual to conclude that the observed 

individual cannot be trusted or relied 

upon. Unlike the background 

investigation and reviews of ‘‘character 

and reputation’’ in § 73.56(d)(6) and 

(k)(1)(v) and proposed § 73.120, which 

are principally retrospective reviews of 

an individual and may be based on 

third-party information (i.e., 

information from individuals not 

subject to NRC requirements), the 

‘‘character or action’’ focus of proposed 

§ 26.606(b)(2)(vi) would be a present 

observation of an individual subject to 

the FFD program and performed by an 

individual who is also subject to the 

FFD program. Whether the information 

would be received from an individual 

subject to the FFD program or someone 

who is not subject to the FFD program, 

the licensee or other entity would need 

to review this information (i.e., 

determine if the information and its 

source are credible) to determine 

whether the individual should maintain 

authorization. 

Proposed § 26.606(b)(3) would require 

licensees and other entities to address in 

their procedures the process, including 

the duties and responsibilities of FFD 

program personnel, to be followed if an 

individual’s behavior or condition raises 

an FFD concern. This provision would 

also require a process to be conducted 

when credible information is received 

by the licensee or other entity that the 

individual is not fit for duty, 

trustworthy, and reliable. 

With a few exceptions, proposed 

§ 26.606(b)(3) would be equivalent to 

current § 26.403(b)(3). Instead of the 

phrase ‘‘while constructing or directing 

the construction of safety- or securityrelated SSCs’’ in current § 26.403(b)(3), 

the NRC would use ‘‘on the NRClicensed facility’’ in proposed 

§ 26.606(b)(3) because this provision 

would apply during commercial nuclear 

plant construction, operation, and 

decommissioning, if applicable, in 

addition to holders of an ML as 

described in § 26.3(f). The requirement 

that the roles and responsibilities of 

FFD program personnel be described 

was developed from current §§ 26.4(g) 

and 26.31(b) and operating experience, 

which has demonstrated that clear job 

descriptions help ensure that 

individuals know who is designated by 

the licensee or other entity to make 

decisions regarding FFD program 

implementation and who can be 

approached when physiological or 

psychological help is needed. This is 

principally a protection consideration 

afforded to individuals subject to the 

FFD program. 

The proposed requirement would also 

include two conditions not found in 

current § 26.403(b) that would clarify 

the initiation of the fitness 

determination process should an 

individual’s behavior or condition raise 

an FFD concern. The phrase, 

‘‘impairment from any cause that in any 

way could adversely affect the 

individual’s ability to safely and 

competently perform the individual’s 

duties,’’ would reflect the § 26.23(b) 

performance objective. The condition, 

‘‘the receipt of credible information 

indicating that the individual cannot be 

trusted or relied on to perform those 

duties and responsibilities making the 

individual subject to this part,’’ would 

reflect the § 26.23(a) performance 

objective. In either case, as required by 

§ 26.23(c), the FFD program must 

provide reasonable measures for the 

early detection of individuals who are 

not fit to perform the duties that require 

them to be subject to the FFD program. 

Proposed § 26.606(b)(4) would require 

licensees and other entities to have 

written procedures that address the 

operation and oversight of an onsite or 

offsite collection facility. This 

requirement would be equivalent to 

current §§ 26.403(b) and 26.405(e) and 

is developed from § 26.41(b), which 

states that each licensee and other entity 

who is subject to subpart B of part 26, 

shall ensure that the entire FFD program 

is audited, which is part of a licensee’s 

or other entity’s oversight of the facility, 

and § 26.87(a), which states that each 

FFD program must have one or more 

designated collection sites that have all 

necessary personnel, materials, 

equipment, facilities, and supervision to 

collect specimens for drug testing and to 

perform alcohol testing. Having 

procedures for the operation and 

oversight of the onsite or offsite 

collection facility would enhance 

consistency in program implementation, 

protect individuals subject to testing, 

and account for the flexibilities afforded 

in the types of biological specimens 

than may be collected under an FFD 

program subject to subpart M of part 26. 

Section 26.606(b)(4), when used with 

the PMRP described in § 26.603(d) and 

the proposed audit requirement in 

§ 26.605(a), would help maintain FFD 

program effectiveness and prevent 

subversion attempts at facilities that 

may not be under the direct day-to-day 

oversight of FFD program personnel. 

Proposed § 26.606(b)(5) would require 

licensees and other entities to have 

written procedures that address the 

fatigue management requirements in 

§ 26.202(b), ‘‘Procedures,’’ and either 

§ 26.205(d)(3) or (d)(7). 

Proposed § 26.606(b)(6) would require 

licensees and other entities to have 

written procedures that provide 

measures to prevent subversion of drug 

and alcohol tests conducted onsite and 

offsite. This proposal was developed 

from § 26.27(c)(1). 

Proposed § 26.607, ‘‘Drug and alcohol 

testing,’’ would establish drug and 

alcohol testing requirements for 

licensees and other entities 

implementing proposed § 26.604, at 

their discretion, and licensees and other 

entities implementing proposed 

§ 26.605. Except for a few differences, 

proposed § 26.607 would be equivalent 

to current § 26.405, which requires 

licensees and other entities 

implementing an FFD program under 

subpart K of part 26 to have a drug and 

alcohol testing program that 

demonstrates compliance with the 

requirements in § 26.405(b) through (g). 

The differences are commensurate with 

the risk consequences presented by a 

part 53-licensed facility as compared to 

a part 50 or 52 nuclear power plant. 

These proposed requirements would 

improve flexibility in the conduct of 


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drug and alcohol testing while 

maintaining protections afforded to 

individuals subject to the FFD program. 

Proposed § 26.607(a) would require 

licensees and other entities to obtain a 

split specimen for all drug tests using 

oral fluid or urine for all test conditions 

in § 26.607(b), (h) and (j). Neither 

current subpart K nor current subparts 

B or E of part 26 require a split 

specimen. However, the majority of the 

LWR fleet uses split specimens for drug 

testing and commercially available drug 

screening products use a split specimen 

technique. Since publication of the 2008 

part 26 final rule, the HHS has issued 

guidelines for urine and oral fluid that 

require split specimens, and the draft 

proposed HHS Guidelines for hair 

requires split specimens, as well. 

The required use of a split specimen 

process would protect the individual 

because, upon a donor-alleged 

discrepant or questionable test result, 

the donor may provide permission to 

test the split specimen (specimen B) in 

an effort to refute the laboratory test 

results for specimen A. The requirement 

also would enable the MRO to direct 

laboratory testing of specimen B if 

specimen A were invalid; though the 

NRC expects specimens becoming 

invalid at the laboratory to be a rare 

occurrence as testing would be 

conducted in HHS-certified laboratories 

with trained collectors. In the event that 

a specimen is determined to be invalid, 

then the occurrence would likely 

warrant further investigation by the 

MRO and laboratory to identify the 

cause. This protocol would be 

equivalent to the special analysis testing 

in current § 26.163(a)(2) for dilute 

specimens in that additional laboratory 

analysis is performed because of a 

questionable test result. 

If a split specimen is tested by an 

HHS-certified laboratory, then the test 

result from specimen B must be used as 

part of the determination for an FFD 

policy violation as required by 

§ 26.185(n), ‘‘Evaluating results from a 

second laboratory.’’ However, this is not 

to say that the test results from 

specimen A should be discarded. Since 

the HHS-certified laboratory should 

report all test results from all specimens 

tested to the MRO, like the information 

described in § 26.169, ‘‘Reporting 

results,’’ test result differences between 

specimens A and B can be used to 

inform the MRO as to what should be 

reported to the licensee or other entity 

to either facilitate medical or clinical 

assistance for the individual, inform an 

FFD-policy violation determination, or 

both. 

The proposed § 26.607(a) requirement 

would also state that if the licensee or 

other entity elects to use a POCTA 

device for screening during random 

testing or portal area monitoring (e.g., 

pre-access screening), a split specimen 

would not need to be taken. The reason 

for this exception would be that the 

requirements in § 26.607(h)(4) establish 

the process to be implemented when a 

screening test indicates a presumptive 

positive, adulterant, or a discrepant 

biological marker, if applicable. This 

process includes collecting and testing a 

specimen for analysis at an HHScertified laboratory. 

Proposed § 26.607(b) would require 

the licensee or other entity to subject 

individuals identified in § 26.202 to 

drug and alcohol testing under the five 

conditions listed in § 26.607(b)(1) 

through (5). Proposed § 26.607(b) would 

be equivalent to current § 26.405(c). 

Proposed § 26.607(b)(1) would require 

pre-access testing similar to current 

§ 26.405(c)(1), which requires testing 

before assignment to construct or direct 

the construction of safety- or securityrelated SSCs. Unlike current 

§ 26.405(c)(1), the proposed requirement 

would not include the phrase, 

‘‘construct or direct the construction of 

safety- or security-related SSCs,’’ 

because, for licensees or other entities 

under part 53, the pre-access test 

condition applies to construction, 

operation, and decommissioning, if 

applicable, to help inform a licensee’s or 

other entity’s authorization 

determination. The proposal also would 

use ‘‘pre-access’’ instead of ‘‘preassignment,’’ which is used in current 

§ 26.405(c)(1). 

A pre-access test would require the 

collection of an oral fluid or a urine 

specimen no more than 14 days before 

the individual is granted unescorted 

access. Although this change has roots 

in the 2008 part 26 final rule, which 

reduced the period within which preaccess testing must be performed from 

60 days to 30 days or less, the 14-day 

proposal is based on three lessons 

learned from operating experience. 

First, the 14-day period would be a 

large enough window of time to collect 

the specimen and evaluate test results 

because licensees or other entities 

typically receive laboratory test results 

within 5 business days of laboratory 

receipt of the biological specimen. At 

the same time, the 14-day period would 

be small enough to help ensure that the 

test results are representative of the 

individual’s forensic toxicology before 

being granted authorization. 

Second, the 14-day window would 

enable the licensee or other entity to 

conduct an unannounced pre-access 

drug and alcohol screening using a hair 

specimen or a POCTA. This would help 

prevent an individual from attempting 

to subvert the drug and alcohol test by 

temporarily abstaining from drug or 

alcohol abuse or adulterating or 

substituting their specimen to obtain a 

non-positive test result. 

Third, the NRC does not expect 

licensees and other entities licensed 

under part 53 to have the large and 

periodic influxes of individuals (either 

licensee employees or C/Vs) that LWRs 

have to support facility operation, 

maintenance, engineering design 

changes, or nuclear refueling. Therefore, 

these licensees or other entities would 

not be periodically challenged to in-take 

a large workforce within the proposed 

14-day pre-access testing window. 

Proposed § 26.607(b)(2) would require 

the licensee or other entity to conduct 

random drug and alcohol testing of all 

individuals subject to the FFD program. 

With one exception, this proposed 

requirement would be equivalent to 

current § 26.405(b). Section 26.405(b) 

gives licensees and other entities that 

implement an FFD program subject to 

subpart K of part 26 the option to 

impose random drug and alcohol 

testing. Proposed § 26.607(b)(2) would 

not offer that option because subpart M 

of part 26, unlike subpart K, would not 

allow a licensee or other entity to 

implement a fitness monitoring program 

under current § 26.406 instead of a 

random testing program. The principal 

reasons for not allowing this flexibility 

would be that no licensee or other entity 

has ever implemented a fitness 

monitoring program (i.e., there is no 

operating or regulatory experience on 

which to judge the effectiveness of a 

fitness monitoring program) and the 

proposed subpart M framework already 

uses behavioral observation to help 

ensure FFD program effectiveness. 

Supplementing the proposed § 26.609 

BOP with an additional observation 

technique (i.e., the fitness monitoring 

program) would not result in a level of 

deterrence or detection equivalent to 

that which would be obtained through 

behavioral observation and random drug 

and alcohol testing. 

Proposed § 26.607(b)(2)(i) through (v) 

would provide specific requirements for 

the conduct of a random testing 

program. These paragraphs would be 

equivalent to § 26.405(b)(1) through (4), 

although with a few differences. The 

similar provisions would be proposed in 

§ 26.607(b)(2)(i), (b)(2)(iii), and 

(b)(2)(iv). 

The differing provisions would 

include proposed § 26.607(b)(2)(ii), 

which would refer to an ‘‘FFD program 

procedure’’ instead of the reference to 

an ‘‘FFD program policy’’ in 

§ 26.405(b)(2) because procedures 


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contain the instructions that implement 

FFD program requirements, but the FFD 

policy need not contain specific 

instructions. Section 26.607(b)(2)(ii) 

would also require individuals who are 

selected for random testing to report to 

the onsite collection site, as opposed to 

the collection site in § 26.405(b)(2) 

because alcohol metabolism necessitates 

a relatively timely alcohol test. This 

change is also proposed because the 

NRC expects that part 53 licensees and 

other entities may use a combination of 

onsite (for random, for-cause, and postevent testing) and offsite (for pre-access, 

post-event, and follow-up testing) 

collection facilities for drug and alcohol 

testing and may have to afford 

reasonable accommodation to certain 

individuals, which would add 

complexity in the licensee’s or other 

entity’s procedurally determined time 

period in which an individual must 

report to the collection facility. 

Another difference from § 26.405(b) 

would be proposed § 26.607(b)(2)(v), 

which would establish the random 

testing rate for the population of 

individuals subject to testing. Subpart K 

of part 26 does not establish a random 

testing rate. The proposed requirement 

would be equivalent to current 

§ 26.31(d)(2)(vii), which requires that 

the sampling process used to select 

individuals for random testing provides 

that the number of random tests 

performed annually is equal to at least 

50 percent of the population that is 

subject to the FFD program. The NRC 

would revise that slightly for proposed 

§ 26.607(b)(2)(v) to require a 50 percent 

random testing rate for the licensee 

employee population and a 50 percent 

random testing rate for the C/V 

population. The NRC proposes this 

change for two reasons. 

First, although operating experience 

has demonstrated that § 26.31(d)(2)(vii) 

helps provide reasonable assurance that 

individuals are fit for duty and 

trustworthy and reliable through the 

detection and deterrence of substance 

abuse, this same operating experience 

demonstrates that, on many occasions, 

the C/V population has been tested at a 

rate lower than 50 percent, even though 

this population results in the majority of 

all FFD policy violations. This bias 

occurs because C/Vs are available for 

testing only during short periods of time 

or periodically throughout the year, 

whereas licensee employees are 

essentially always available for a test. 

A second reason why the NRC is 

proposing a different 50 percent random 

testing protocol than in the current part 

26 requirements is that the flexibilities 

afforded to part 53 licensees or other 

entities in subpart M of part 26 are not 

afforded to licensees or other entities 

that must implement an FFD program 

under subparts A through I, N, and O of 

part 26. These flexibilities include 

enabling the use of a POCTA device to 

screen individuals during the random 

testing process and the use of offsite 

collection facilities for pre-access 

testing. The potential reduction in FFD 

program effectiveness caused by 

licensee or other entity implementation 

of these options would be offset by 

subpart M requirements that mitigate 

possible challenges to the FFD program, 

such as the 50 percent random testing 

rate for the licensee employee 

population and 50 percent random 

testing rate for the C/V population. 

Proposed § 26.607(b)(3) would require 

for-cause testing equivalent to that used 

in current FFD programs implementing 

§ 26.405(c)(2). The NRC would require 

for-cause testing, like random testing, to 

be conducted onsite to ensure that the 

test is conducted as soon as reasonably 

practicable. This is an important 

consideration when for-cause testing for 

alcohol or using oral fluid for drug 

screening or testing because human 

metabolism continually lowers the 

concentrations of the drugs, drug 

metabolites, and alcohol perhaps to 

concentrations lower than the initial or 

confirmatory testing cutoffs. 

Additionally, for facilities that are sited 

in geographically remote locations, an 

offsite collection facility might be too far 

away or not readily accessible. 

Proposed § 26.607(b)(4) would require 

post-event testing in a manner 

equivalent to current § 26.405(c)(3) with 

a few adjustments. For part 53 licensees 

or other entities, the NRC proposes postevent testing under two conditions: 

events involving human errors that may 

have caused or contributed to the events 

(proposed § 26.607(b)(4)(i)), and events 

not involving human error that result in 

adverse health consequences or damage 

to any safety- or security-related SSC 

(proposed § 26.607(b)(4)(ii)). The word 

‘‘significant’’ would not be used in 

§ 26.607(b)(4)(ii)(A) to describe the 

‘‘illness or personal injury’’ as used in 

§ 26.405(c)(3)(i) because 

§ 26.607(b)(4)(ii)(A) would describe 

which illnesses or injuries are covered. 

Proposed § 26.607(b)(4)(ii)(B), unlike 

§ 26.405(c)(3)(ii), would not use the 

word ‘‘significant’’ to describe the 

damage to safety- or security-related 

SSCs because any damage to safety- or 

security-related SSCs would require 

testing within four hours of the event 

unless immediate medical intervention 

precludes the conduct of the test on the 

individual(s) who caused or contributed 

to the event. Proposed 

§ 26.607(b)(4)(ii)(B) also would not use 

the word ‘‘construction’’ as in 

§ 26.405(c)(3)(ii) because § 26.607(b)(4) 

would apply to construction, operation, 

and decommissioning, if applicable. 

Proposed § 26.607(b)(4)(i) would 

require the licensee or other entity to 

define in its procedures the terms 

‘‘human error’’ and ‘‘event.’’ These 

terms may take on various meanings 

and they are not defined in the current 

or proposed rule, so the licensee or 

other entity would be required to 

describe or define these terms to help 

ensure consistent implementation of 

subpart M of part 26 and that the postevent test condition would be 

consistently applied to all individuals 

subject to the FFD program. The 

§ 26.405(c)(3)(i) requirement that ‘‘the 

event is recordable under the 

Department of Labor standards 

contained in 29 CFR 1904.7, and 

subsequent amendments thereto,’’ 

would not be carried over to proposed 

§ 26.607(b)(4). Instead, the NRC 

proposes to prescribe the post-event test 

conditions in § 26.607(b)(4), in part so 

they would not change unless the NRC 

amends the requirement. 

Proposed § 26.607(b)(5) would require 

follow-up testing. This requirement 

would be equivalent to current 

§ 26.405(c)(4), although the proposed 

§ 26.607(b)(5) would further describe 

follow-up testing. The NRC proposes to 

describe follow-up testing as part of a 

series of tests for drugs, alcohol, or both, 

which are performed after an individual 

subject to part 26 has violated the FFD 

policy on substance use or abuse, or the 

sale, use, or possession of illegal drugs. 

Follow-up testing would be used to 

verify an individual’s continued 

abstinence from substance abuse. The 

NRC would not include a reference to a 

follow-up plan as in § 26.405(c)(4) 

because the intent of a follow-up plan 

is to conduct a series of drug tests, 

alcohol tests, or both, to verify 

continuing abstinence from substance 

abuse. Nevertheless, individuals who 

violate an FFD policy on substance use 

or abuse, or the sale, use, or possession 

of illegal drugs, should have a follow-up 

plan that includes a definition of 

‘‘abstinence’’ from the medical 

professional prescribing the plan. 

Proposed § 26.607(c) would provide 

additional testing requirements. This 

proposed requirement would be 

equivalent to § 26.405(d) and would 

require implementation of select 

requirements from current subpart E of 

part 26. The proposed requirements 

would govern directly observed 

collections, shy bladder situations, 

special analysis testing, and alcohol 

testing. These requirements would be 

necessary to maintain FFD program 


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effectiveness equivalent to that 

currently implemented by the LWR 

fleet. 

Proposed § 26.607(c)(1) would require 

validity testing and establish the 

minimum panel of drugs and drug 

metabolites to be tested. This panel 

would be the same as those in 

§§ 26.31(d)(1) and 26.405(d) because, 

based on operating experience from 

LWR FFD program implementation, this 

panel has been determined to contribute 

to a licensee or other entity satisfying 

the FFD performance objectives in 

§ 26.23(a) through (d). 

Proposed § 26.607(c)(1) would differ 

from § 26.405(d) because it would 

require testing of oral fluid and urine 

specimens for validity, including at 

least one biological marker (developed 

from an HHS Guidelines provision) and 

one adulterant (equivalent to current 

validity testing for urine specimens in 

part 26). Section 26.405(d) requires that 

urine specimens collected for drug 

testing be subject to validity testing. The 

addition of oral fluid validity testing is 

important because, just as there are 

publicly available kits to subvert a urine 

drug test, kits that may be used to 

subvert a drug test that uses oral fluid 

as a biological specimen are also readily 

available. 

Proposed § 26.607(c)(2) would 

include requirements that already exist 

in the part 26 framework that provide 

protections for individuals subject to the 

FFD program and contribute to testing 

effectiveness when collecting and 

assessing a urine specimen. Specifically, 

current § 26.115, ‘‘Collecting a urine 

specimen under direct observation,’’ 

describes the exclusive grounds for 

performing a directly observed 

collection and the process to be 

followed to protect the privacy of the 

individual. Section 26.119, 

‘‘Determining ‘shy’ bladder,’’ establishes 

the process to be followed when a donor 

is not able to produce a sufficient 

amount of urine for testing, and 

§ 26.163(a)(2) requires special analysis 

testing when a specimen is dilute to 

help prevent a subversion attempt. 

Proposed § 26.607(c)(3) would require 

implementation of all the current 

alcohol testing requirements in § 26.91, 

‘‘Acceptable devices for conducting 

initial and confirmatory tests for alcohol 

and methods of use,’’ through § 26.103, 

‘‘Determining a confirmed positive test 

result for alcohol.’’ Using the same 

alcohol testing framework for parts 50, 

52, 70, and 53 licensees and other 

entities would provide for regulatory 

consistency, protections for individuals 

subject to the FFD program (e.g., the 

quality controls and verification applied 

to the EBT device), and FFD program 

effectiveness (e.g., accuracy of test 

results). For alcohol testing, unlike drug 

testing, there is a preponderance of 

evidence that correlates blood alcohol 

concentrations to impairment and 

intoxication. Furthermore, FFD 

performance data has demonstrated that 

the time-dependent alcohol cutoffs in 

§ 26.103 have increased the detection of 

individuals who are under the influence 

of alcohol. For these reasons, the current 

alcohol requirements in part 26 are 

proposed for FFD programs under 

subpart M. 

Proposed § 26.607(c)(4) would 

establish additional testing 

requirements. This proposal would be 

equivalent to current § 26.405(f) for 

facilities licensed under part 53 for the 

conduct of drug testing. Unlike 

§ 26.405(f), proposed § 26.607(c)(4) 

would not reference validity screening 

and initial drug and validity tests at 

licensee testing facilities as this would 

be required in proposed § 26.607(c)(1). 

Another minor difference between 

§ 26.405(f) and proposed § 26.607(c)(4) 

would reflect the requirement in subpart 

M of part 26 to use an HHS-certified 

laboratory for all biological specimens 

collected and not just for urine 

specimens. 

Consistent with § 26.405(f), proposed 

§ 26.607(c)(4) would require the use of 

an HHS-certified laboratory for all test 

conditions listed in § 26.607(b), MROdirected tests, and the testing of a split 

specimen. Further, HHS-certified 

laboratory test results using urine or oral 

fluid would be required for the issuance 

of an FFD policy violation and part 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 

§ 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 §§ 26.607(i)(3) and 26.610 for 

subversions, as defined in § 26.5, of the 

drug and alcohol screening process. 

There are three phrases or 

requirements in § 26.405(f) that the NRC 

does not propose to use in 

§ 26.607(c)(4). The first is the phrase, 

‘‘consistent with its standards and 

procedures for certification,’’ regarding 

the operation of an HHS-certified 

laboratory, because the laboratory 

would not be HHS-certified if it were 

not following ‘‘its standards and 

procedures for certification.’’ The 

second is the requirement that urine 

specimens that yield positive, 

adulterated, substituted, or invalid 

initial validity or drug test results must 

be subject to confirmatory testing by the 

HHS-certified laboratory, except for 

invalid specimens that cannot be tested. 

This requirement would not be used 

because, under subpart M of part 26, 

licensees or other entities would be 

required to use an HHS-certified 

laboratory. For a laboratory to be HHScertified, it must follow the HHS 

Guidelines and include procedures that 

describe when a specimen cannot be 

tested. Lastly, the § 26.405(f) 

requirement that other specimens that 

yield positive initial drug test results 

must be subject to confirmatory testing 

by a laboratory that demonstrates 

compliance with stringent quality 

control requirements that are 

comparable to those required for 

certification by the HHS, would not be 

used because subpart M of part 26 

would require the use of an HHScertified laboratory. 

Proposed § 26.607(c)(4) would require 

the licensee or other entity to contract 

with a primary and backup HHScertified laboratory. This provision 

would help ensure that specimens are 

processed and tested to maintain FFD 

program effectiveness should the 

primary laboratory be unable to perform 

specimen testing. This would help 

maintain protections afforded to 

individuals subject to the FFD program 

(e.g., should the donor or MRO request 

testing of the split specimen, a different 

laboratory could be used). This 

requirement also would state that the 

primary and backup laboratories must 

have a different certifying scientist. 

Having a back-up HHS-certified 

laboratory and a different certifying 

scientist would benefit the program and 

donor because the drug testing 

instruments, technicians, and certifying 

scientist would be independent of the 


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primary laboratory testing and review 

process. The back-up HHS-certified 

laboratory may be of the same corporate 

entity as the primary laboratory. 

Proposed § 26.607(c)(4) would also 

state that the laboratory would be 

subject to inspection or audit by the 

licensee or other entity and that records 

and documents must be provided and/ 

or able to be photocopied and removed 

from the premises to support the 

inspection or audit. This requirement 

would be equivalent to current 

§ 26.41(d) except that laboratories 

would not be able to limit the use and 

dissemination of documents copied or 

taken from the laboratory by a licensee 

or other entity. This is necessary to 

ensure the continuing effectiveness of 

FFD programs, because NLCP findings 

and audit results could adversely 

impact FFD program effectiveness. 

Pertinent information includes and 

should not be limited to NLCPidentified weaknesses (e.g., custody and 

control, accessioning, instrumentation, 

procedures, training, supervision, 

review of test results, and resolution of 

previously identified corrective actions) 

that may impact the effectiveness of 

FFD programs. 

Proposed § 26.607(d) would help 

protect the donor from mistakes made 

during the drug and alcohol testing 

processes and help ensure FFD program 

effectiveness. The rule would require 

the licensee or other entity to protect the 

individual’s privacy and the integrity of 

the specimen and to implement quality 

controls to ensure that test results are 

valid and attributable to the correct 

individual. This requirement would be 

equivalent to the first sentence of 

current § 26.405(e), except that the word 

‘‘stringent’’ was removed from the 

phrase ‘‘stringent quality controls,’’ 

because the word ‘‘stringent’’ is not 

defined. 

Proposed § 26.607(e) would describe 

the requirements for licensees and other 

entities that use offsite collection 

facilities. Consistent with current 

§ 26.405(e), a licensee or other entity 

would be able to conduct specimen 

collections and alcohol testing at a local 

hospital or other facility. Unlike 

§ 26.405(e), proposed § 26.607(e) would 

not restrict licensees and other entities 

to use hospitals and other facilities that 

meet the requirements in 49 CFR part 

40, ‘‘Procedures for Transportation 

Workplace Drug and Alcohol Testing 

Programs,’’ because subpart M of part 26 

is intended to provide flexibilities 

beyond those in the current part 26 

framework. Licensees and other entities 

may use these Department of 

Transportation requirements to inform 

their procedures under § 26.606(b)(1) as 

long as the procedures do not conflict 

with the requirements in part 26 or the 

HHS Guidelines. 

Proposed § 26.607(e) would also 

require licensees and other entities to 

audit offsite collection facilities before 

their use and biennially to confirm that 

the facility procedures are comparable 

to those described in subpart E of part 

26 or the HHS Guidelines for urine and 

oral fluid. This proposed requirement is 

based on current § 26.41(a) and (b). The 

§ 26.607(e) audit requirement would be 

a program effectiveness consideration 

because offsite collection facilities may 

not require vigilance of their collectors 

(e.g., identification of subversion 

attempts), diligence in the protection of 

worker rights (e.g., privacy and 

specimen custody and control), or 

procedural compliance. 

The offsite facility used by a licensee 

or other entity under proposed 

§ 26.607(e) would have to be licensed to 

conduct specimen collections and 

perform alcohol testing, and be audited, 

by the State or a State-designated entity. 

This requirement would help provide 

assurance of adequate collection facility 

performance and may help reduce the 

burden on the licensee or other entity 

and the collection facility. Crediting a 

State audit (or State licensure, oversight, 

or regulation) is established in 

§§ 26.4(i)(4) and (j), 26.91(e)(5), 

26.153(f)(1), and 26.183(a). 

Proposed § 26.607(f) would provide 

the requirements for initial drug testing. 

This provision would be equivalent to 

§ 26.405(f) except to account for the 

alternative biological specimens that 

may be tested under subpart M of part 

26. For the testing of all biological 

specimens, the licensee or other entity 

under part 53 would be required to use 

a device that employs an immunoassay 

screening technique, or an alternative 

technology that the licensee or other 

entity has incorporated into its FFD 

program through the § 26.603(e) change 

control process, that demonstrates 

compliance with the requirements of the 

U.S. Food and Drug Administration 

(FDA) for commercial distribution. 

Examples of alternative technologies 

include liquid or gas chromatography 

and mass spectrometry. Licensees and 

other entities would use the § 26.603(e) 

change control process to evaluate and 

document a change to their collection 

and analysis procedures to enable the 

use of a better or perhaps more costeffective collection and/or testing 

technology. Another difference from 

§ 26.405(f) would be changing the word 

‘‘urine’’ in § 26.405(f) to ‘‘biological 

specimens’’ in § 26.607(f). Lastly, 

proposed § 26.607(f) would include the 

phrase ‘‘discrepant biological marker’’ 

as a drug screening result that must be 

analyzed by an HHS-certified laboratory 

and evaluated by the MRO to help 

inform the MRO’s determination of a 

subversion attempt. 

Proposed § 26.607(g) would enable a 

part 53 licensee to use oral fluid as a 

biological specimen for testing. This 

requirement would be equivalent to 

§ 26.31(d)(5), which enables the MRO to 

conduct drug and alcohol testing using 

alternative methods, and § 26.405, 

which does not preclude the use of oral 

fluid specimens for FFD programs that 

implement subpart K of part 26 

requirements. In order to provide 

assurance that drug testing is effective 

and protects the worker, § 26.607(g) 

would require that the licensee’s or 

other entity’s procedures incorporate 

the HHS Guidelines or the requirements 

in part 26 for the conduct of urine or 

oral fluid testing. 

The proposed § 26.607(g) requires that 

the oral fluid collection device must 

have received premarket approval from 

the FDA and must not expire before 

laboratory testing. Also, the drugs, drug 

metabolites, initial and confirmatory 

testing cutoffs, and biological markers, if 

applicable, must be those established by 

HHS for oral fluid drug testing and the 

alcohol cutoffs in part 26. If they are not 

established by HHS or the NRC for the 

paneled drugs and drug metabolites, 

then they would be determined and 

documented by a forensic toxicologist 

review. This forensic toxicologist review 

would help ensure that the device 

accurately tests for the drug, drug 

metabolite, biological markers, 

adulterants, and/or alcohol and that the 

results from the device are comparable 

to those established in the HHS 

Guidelines for oral fluid testing. 

Proposed § 26.607(h)(1) and (2) would 

enable the use of a POCTA device 

during the random and pre-access 

testing processes. These requirements 

are adopted from § 26.97, ‘‘Collecting 

oral fluid specimens for alcohol and 

drug testing,’’ and § 26.405(f), which 

does not preclude the use of oral fluid 

testing. To use a POCTA device for 

urine, oral fluid, or other biological 

indicators (breath, sweat, etc.), a 

forensic toxicology review would be 

required to ensure that the device is 

forensically effective. If the POCTA 

device is forensically effective, then the 

donor would be reasonably protected 

from a false positive test result, the 

licensee or other entity would be 

reasonably protected from false negative 

test results, and the FFD program would 

remain effective. For a POCTA device to 

be forensically effective, the forensic 

toxicologist would need to document an 

evaluation that the performance of the 


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POCTA device must be comparable to 

the requirements in § 26.161(b) for a 

urine specimen or the procedures in the 

HHS Guidelines for urine or oral fluid, 

as implemented by the licensee or other 

entity through its procedures. 

The use of POCTA for oral fluid and 

urine specimens for the pre-access and 

random testing processes would be 

acceptable because individuals in the 

pre-access process would be subject to 

an oral fluid or urine specimen 

collection and possible drug screening 

using a hair specimen, which are both 

required to be sent to an HHS-certified 

laboratory. For random testing, the 

individual would have also been 

granted authorization under the AA and 

FFD requirements and have been subject 

to behavioral observation and physical 

protection screening (e.g., verification of 

identify, and screening for explosives 

and contraband). 

Proposed § 26.607(h)(3) would require 

that procedures be developed that 

ensure the effectiveness of the POCTA 

collection process, assessment of the 

screening results, and prevention of 

subversion attempts. This requirement 

would be equivalent to current 

§ 26.403(b)(1) and would help ensure 

protections afforded to individuals 

subject to the FFD program and program 

effectiveness. The subpart M of part 26 

framework enables the use of POCTA 

for random screening of individuals for 

any part 53 facility, so the licensee or 

other entity should exercise due 

diligence and implement risk 

management strategies to ensure the 

efficacy of random screening and its 

contribution to an effective FFD 

program. 

Proposed § 26.607(h)(4) would 

provide that an individual donor who 

screens positive (or whose specimen is 

invalid or indicates a discrepant 

biological marker or adulterant) is 

removed from all duties and 

responsibilities making the donor 

subject to subpart M of part 26. Under 

proposed § 26.607(h)(4)(i), the donor 

then would be immediately subject to a 

drug and alcohol test that provides 

quantified confirmatory test results from 

which an FFD policy violation may be 

issued. Similar to other requirements for 

specimen collections, except for 

biological specimens analyzed by a 

passive detection system, the licensee or 

other entity would be required to 

implement procedures that ensure that 

all specimens collected are uniquely 

assigned to the donor (i.e., procedures 

that provide for custody and control of 

the specimen). If the individual shows 

signs of impairment during the POCTA 

process, proposed § 26.607(h)(4)(ii) 

would require the temporary removal of 

the individual’s authorization until the 

MRO reviews the laboratory test 

result(s), and interviews the individual, 

and a determination of fitness finds that 

authorization may be restored. Section 

26.607(h)(4) is equivalent to § 26.77(b) 

and was informed by the requirements 

in §§ 26.419, 26.75(c) and (d), and 

26.185(c). 

Proposed § 26.607(i) would enable the 

collection of hair specimens for drug 

testing to supplement pre-access testing 

that uses urine or oral fluid specimens. 

Hair testing would be a new feature in 

the part 26 framework. The NRC 

proposes to permit the use of hair 

testing for only Schedule I or II drugs or 

their metabolites to inform a licensee’s 

or other entity’s determination whether 

the individual is trustworthy and 

reliable. For example, if an individual 

stated no prior use of illegal drugs or 

potentially addictive habits, a hair 

screening test could be performed 

during the pre-access process to 

ascertain the validity of the individual’s 

statement. However, if the HHS-certified 

laboratory communicates a laboratoryconfirmed positive test result, an FFD 

policy violation may not be 

administered. This laboratory 

information must be treated as 

potentially disqualifying FFD 

information, unless the individual 

subverts the screening process, in which 

case a permanent denial of 

authorization must be issued under 

proposed § 26.610. To provide 

assurance of testing effectiveness and 

protections afforded to individuals 

subject to the FFD program, proposed 

§ 26.607(i) would require that an HHScertified laboratory must be used to 

analyze the hair specimen, a forensic 

toxicologist must review the licensee’s 

or other entity’s hair screening process, 

the test kit must be cleared by the FDA, 

and hair screening must be conducted 

in accordance with the HHS Guidelines. 

The forensic toxicologist review would 

be necessary if the panel of drug or drug 

metabolites to be tested and their cutoffs 

are not established by HHS or the NRC 

for hair. 

Proposed § 26.607(j) would allow the 

use of portal area screening for drugs, 

alcohol, or both. This provision would 

result in a substantial contribution to a 

licensee or other entity satisfying the 

§ 26.23 performance objectives by 

helping ensure that 100 percent of all 

individuals who arrive at the NRClicensed facility to perform or direct 

those duties and responsibilities or 

maintain those types of access making 

them subject to the FFD program are fit 

for duty and deterred from arriving 

onsite in a physiological condition that 

may be adverse to safety and security. 

Additionally, screening could be 

conducted when an individual exits the 

NRC-licensed facility to provide 

assurance that substance abuse had not 

occurred on the site (see § 26.23(d)). The 

screening device could be electronically 

linked to temporarily prevent ingress or 

egress and could automatically inform 

licensee- or other entity-designated 

officials of the portal area alarm. The 

proposed requirement would enable the 

licensee or other entity to use 

innovative technologies to maintain 

FFD program effectiveness when their 

PMRP compels the licensee or other 

entity to implement mitigative strategies 

to maintain program effectiveness. The 

use of portal screening technologies may 

also represent cost savings because, for 

NRC-licensed facilities that have small 

staff sizes or are geographically remote, 

passive drug and alcohol screening 

technologies could be an innovative 

alternative to a random testing program, 

although the license or other entity 

would need to request and receive an 

exemption. 

Proposed § 26.607(j) would also 

provide that if the portal area screening 

instrument detects a substance that 

exceeds the instrument’s established 

setpoint, the individual would be tested 

with either a collection kit that must be 

analyzed by an HHS-certified laboratory 

or a POCTA. This situational screening 

would be equivalent to a for-cause test. 

The requirements would not allow an 

individual to be rescreened by the portal 

area screening instrument following an 

initial screening detection that exceeded 

an established setpoint in order to 

prevent a subversion attempt. Similar to 

other drug and alcohol testing 

technologies enabled for use by subpart 

M of part 26, a forensic toxicology 

review would be required before using 

passive screening technology to help 

ensure the effectiveness of the 

instrument by protecting against false 

positive or negative screening results, 

which would place an unwarranted 

burden on the individual, licensee, or 

other entity. These instruments and 

alcohol screening devices, already in the 

marketplace, may also be used to 

determine true identity to facilitate 

implementation of the FFD BOP, which 

may be very practicable at facilities that 

operate with small staff sizes. 

Proposed § 26.607(k) would enable 

the use of a blood specimen for drug, 

alcohol, or other testing for certain 

medical conditions as determined by 

the licensee- or other entity-designated 

MRO. This requirement would be 

equivalent to current § 26.31(d)(5). The 

use of a licensee- or other entitydesignated MRO and not one designated 

by a third party, such as an MRO 


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employed by an offsite specimen 

collection facility, is important because 

the MRO must be familiar with the 

subpart M of part 26 requirements. To 

help ensure testing effectiveness and 

protect the worker, the blood test would 

need to be conducted by a laboratory 

that demonstrates compliance with 

quality control requirements that are 

comparable to those required for 

certification by the HHS, such as a 

hospital or clinic certified by the State, 

Commonwealth, or territory. 

Proposed § 26.607(l) would require 

licensee and other entities to use a 

Federal custody-and-control form (CCF) 

approved by the OMB for the collection 

and packaging of a hair, oral fluid, or 

urine specimen. This proposed 

requirement is based on the CCF 

documentation requirements in current 

subpart E of part 26 because subpart K 

of part 26 does not require the use of a 

CCF under § 26.117(e). Additionally, 

when using a POCTA device, the 

licensee or other entity would be 

required to implement a licensee- or 

other entity-approved and -maintained 

procedure that ensures the reliability of 

the tracking, handling, and storage of a 

specimen from the point of specimen 

collection to final disposition of the 

specimen and the reliability of an 

identification system to uniquely assign 

the specimen to the donor. Both 

requirements would help protect the 

worker by helping ensure chain of 

custody and by contributing to program 

effectiveness. 

Proposed § 26.607(m) would establish 

requirements for the licensee- or other 

entity-designated MRO. Section 

26.607(m)(1) would be equivalent to 

§ 26.405(g), however, the word 

‘‘designated’’ would be added to the 

first sentence to clarify that the MRO 

would be designated by the licensee or 

other entity, and not by a third party. As 

stated with regard to proposed 

§ 26.607(k), this change would clarify 

that it is the licensee’s or other entity’s 

responsibility, through their designated 

MRO, to determine whether an 

individual is fit for duty and 

trustworthy and reliable. This would be 

consistent with the description of FFD 

program personnel in current § 26.31(b) 

and help provide FFD program 

effectiveness and protections to 

individuals subject to the FFD program. 

The paragraph was also modified from 

§ 26.405(g) to address the 

determinations of FFD policy violations 

and fitness required by subpart H for a 

part 53 licensee or other entity that 

implements the FFD program described 

in § 26.605(b). 

Proposed § 26.607(m)(2) would help 

ensure that MRO reviews are consistent 

with those MRO reviews conducted at 

other NRC-licensed facilities subject to 

part 26 and that the MRO maintains 

knowledge of drug collection, testing 

processes and procedures, and 

evaluation of testing results. 

The NRC also proposes that if an 

MRO performed the duties and 

responsibilities in §§ 26.185 and 26.187 

for at least three continuous years in the 

last 10 years prior to being hired or 

contracted by the licensee or other 

entity, then the MRO would not need to 

repeat the initial training and 

examination requirements. The basis for 

3 years is that the MRO would have 

experienced three annual cycles of 

evaluating drug and alcohol test results, 

contributed to the FFD annual report to 

the NRC, experienced a refueling or 

maintenance outage, understood the 

duties and responsibilities of 

individuals subject to the FFD program 

to make informed determinations of 

fitness, demonstrated a safety culture 

that helps ensure FFD program 

effectiveness, and been subject to NRC 

inspection. The basis for 10 years is the 

relatively long periods between 

significant changes to part 26 and the 

HHS Guidelines. 

Proposed § 26.607(m)(3) would 

require that the MRO attend a medicalor clinical-based training session on a 

triennial basis. This proposal was 

developed from Section 13.1 of the HHS 

Guidelines for urine and oral fluid with 

two substantial differences: the HHS 

Guidelines state that ‘‘requalification 

training,’’ including an exam, must be 

conducted ‘‘at least every 5 years from 

initial certification,’’ whereas the 

proposed § 26.607(m)(3) would require a 

training session every three years. The 

proposed requirements are justified 

because changes in societal drug use or 

forensic toxicology could occur more 

frequently than every 5 years, which 

could compel MROs to attend training 

in areas of forensic toxicology, 

determinations of fitness, or other part 

26 technical areas on a more frequent 

periodicity than every 5 years to 

improve their knowledge and expertise. 

Proposed § 26.607(m)(4) would 

require the MRO to evaluate drug testing 

results by implementing the 

requirements in § 26.185 or the HHS 

Guidelines through the licensee’s or 

other entity’s procedures. This 

requirement would help ensure FFD 

program effectiveness and enhance 

consistency across the commercial 

nuclear industry for the evaluation of 

drug testing results. This also would 

help protect individuals because they 

would be subject to the same evaluation 

criteria. If § 26.185 provides insufficient 

information for an MRO to make a 

determination on a drug testing result 

(including adulterant and discrepant 

biological markers), the guidance issued 

by a State agency in the state in which 

the NRC-licensed facility is located, 

Federal agency, or nationally recognized 

MRO training and certification 

organization may be used to inform an 

MRO determination. This provision 

would ensure that the MRO has the 

flexibility to inform their evaluation of 

the drug testing results and fitness 

determination, if necessary, considering 

the drug- and alcohol-related 

flexibilities afforded in subpart M of 

part 26. 

The proposed requirement would also 

state that an MRO need not review a 

confirmed alcohol positive test result 

determined by an EBT device under 

§ 26.607(c)(3)(vi) and (vii), which are 

equivalent to the current requirements 

in §§ 26.101 and 26.103, respectively. 

The results of an EBT device are precise 

and accurate enough to support the 

issuance of an FFD policy violation 

without an MRO review of an EBT test 

result if the instrument demonstrates 

compliance with the requirements in 

§ 26.91. The NRC acknowledges that 

there are physiological conditions that 

may cause an abnormally high blood 

alcohol concentration, such as diabetes, 

acid reflux, gastroesophageal reflux 

disease, and perhaps certain diets (high 

protein and low carbohydrates). 

However, operating experience has not 

demonstrated a compelling need to 

require an MRO review of all EBT test 

results. For consistency, a licensee or 

other entity may elect to require its 

MRO to review all EBT test results when 

a donor communicates a testing concern 

or physiological condition. If the donor 

has a testing concern, the occurrence 

could be appealed under the proposed 

§ 26.613. If the donor presents a 

physical condition to the MRO that may 

have caused an elevated EBT test result, 

the MRO may direct an alternative 

testing process (see § 26.607(m)(5)) 

should it be medically necessary. 

Proposed § 26.607(m)(5) would 

require the licensee- or other entitydesignated MRO to determine and 

approve the use of oral fluid or urine as 

an alternative biological specimen when 

the donor cannot provide a requested 

specimen for testing. This proposed 

requirement is equivalent to 

§ 26.31(d)(5), which enables the use of 

an alternative specimen collection if a 

medical condition makes the collection 

of the biological specimen difficult. This 

determination and the retest must be 

completed as soon as reasonably 

practicable and documented to support 

recordkeeping, auditing, and NRC 

inspection. 


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Proposed § 26.607(m)(6) would 

require that the MRO review all 

specimens screened or tested associated 

with a drug-related FFD policy 

violation. This includes POCTA, split 

specimens, and all specimens taken to 

resolve a discrepant condition, such as 

a possible subversion attempt, 

impairment without a known cause, or 

a donor-requested or MRO-directed 

retest. To resolve a discrepant 

condition, the MRO is authorized to test 

a specimen for a biological marker, 

adulterants, or additional drugs. The 

broad scope of this MRO evaluation 

would be necessary because of the 

variety of different screening and testing 

methods that may have been associated 

with the FFD policy violation. All 

information learned from the conduct of 

part 26 drug and alcohol screening and 

testing should be used in the evaluation 

of an individual’s trustworthiness and 

reliability, issuance of a sanction, and 

development of a follow-up treatment 

and testing plan, if administered. 

Proposed § 26.607(n) is equivalent to 

current § 26.31(d)(6) and would 

establish limits on the screening and 

testing of biological specimens. This is 

a protection consideration afforded to 

individuals subject to the FFD program 

and was not provided in subpart K of 

part 26. This requirement states that 

specimens collected under NRC 

regulations may only be designated or 

approved for screening and testing as 

described in this part and may not be 

used to conduct any other analysis or 

test without the written permission of 

the donor. Analyses and tests that may 

not be conducted include, but are not 

limited to, deoxyribonucleic acid (i.e., 

DNA) testing, serological typing, or any 

other medical or genetic test used for 

diagnostic or specimen identification 

purposes. 

The NRC proposes to require that no 

biological specimens may be passively 

sampled and analyzed in a manner 

different than described in subpart M of 

part 26 to ensure workers are protected 

from non-consensual passive screening. 

The subpart M framework enables 

passive detection of drugs and alcohol, 

whereas passive detection is not 

afforded in subparts A through I, N, and 

O of part 26. 

Proposed § 26.607(o) is equivalent to 

current §§ 26.31(b)(1)(iii)(A) and 26.89 

and would require that all specimen 

collections be conducted by a licenseeor other entity-designated and -trained 

individual. For subpart M of part 26, 

this would include onsite specimen 

collections, except a collection by a 

portal area screening instrument in 

§ 26.607(j). 

Proposed § 26.608 would require 

licensees and other entities to provide 

FFD program training to individuals 

subject to the FFD program. The 

proposed performance-based § 26.608 

requirement was developed from the 

prescriptive training requirements in 

current § 26.29 and modeled on current 

§ 50.120 and the proposed requirements 

in §§ 53.725 and 53.830 because there is 

no training requirement in subpart K of 

part 26. 

Proposed § 26.608(a)(1) would require 

an FFD training program that includes 

the licensee’s or other entity’s FFD 

policies and procedures, including 

fatigue management, and the 

individuals’ FFD program 

responsibilities. Individuals who collect 

specimens for testing or screening must 

also be trained in specimen collector 

duties and responsibilities, including, at 

a minimum, specimen collection, 

custody and control, identification and 

response to subversion attempts, and 

privacy. The fatigue management 

training must include the knowledge 

and abilities described in § 26.202(c). 

For individuals specified in § 26.4, a 

licensee or other entity of a commercial 

nuclear plant would be required to use 

a SAT as defined in proposed in 

§ 53.725. These requirements are based 

on requirements in § 26.29(a)(2), (3), (9), 

and (10). 

Proposed § 26.608(a)(2) would require 

training on the BOP. This requirement 

would be based on §§ 26.29(a)(8), (9), 

and (10) and 26.33. The proposal would 

require individuals to be trained in the 

detection of behaviors or conditions 

related to not only illegal drugs, as in 

the current § 26.33 BOP requirements, 

but also illicit drugs and substance 

abuse onsite and offsite. Also, in 

reference to impairment from fatigue or 

any cause if left unattended, the phrase 

in § 26.33, ‘‘may constitute a risk to 

public health and safety or the common 

defense and security,’’ would be 

replaced in § 26.608(a)(2)(iii) with 

‘‘could result in inattentiveness or 

human errors,’’ because subpart M of 

part 26 is focused, in part, on ensuring 

individuals are fit for duty to safely and 

competently perform or direct the 

performance of assigned duties and 

responsibilities. 

Proposed § 26.608(a)(2)(iv) focuses on 

training to inform individuals that they 

are responsible for their own conduct, 

as well as observing others. Specifically, 

individuals would be trained to 

recognize when they feel unable to 

safely and competently perform 

assigned duties and responsibilities or 

act in a trustworthy and reliable 

manner. The proposed training 

requirement and the proposed reporting 

requirement in § 26.606(a)(5) are in the 

interest of safety and security because 

the individual is proactively 

announcing that assistance may be 

necessary. This would be consistent 

with the performance objectives in 

§ 26.23(b) and (c) where certain 

behavior or stress conditions may be 

indicative of an individual not being fit 

for duty, trustworthy, and reliable. 

Proposed § 26.608(a)(3) would help 

ensure that individuals subject to the 

FFD program understand that FFD 

policy violations would result in an FFD 

program sanction and that program 

information learned or generated by 

FFD program implementation would be 

used to aide licensee or other entity 

authorization determinations and be 

shared, as requested, with other 

licensees or other entities subject to 

parts 26, 53, and 73. This proposed 

requirement is equivalent to 

§ 26.29(a)(1). Proposed § 26.608(a)(3) 

would be a protection measure afforded 

to individuals subject to the FFD 

program because they would 

understand that licensees and other 

entities subject to parts 26, 53, and 73 

would be informed of, in part, an 

individual’s character, reputation, and 

ability to follow policies, procedures, 

and instructions to safely and 

competently perform assigned duties 

and responsibilities in a trustworthy 

and reliable manner. Fitness-for-dutyrelated information would include drug 

and alcohol testing results (not 

quantitative testing values), issuance of 

any sanctions, FFD-determinations 

regarding trustworthiness and 

reliability, testing programs, treatment, 

and other remedial or corrective action. 

Proposed § 26.608(b) would require 

individuals be trained and receive a 

trainee assessment before pre-access 

testing and that refresher training and 

trainee assessments be conducted 

periodically thereafter. These 

requirements would be equivalent to 

§ 26.29(c)(1). However, § 26.608(b) was 

developed from the SAT-based training 

requirements in § 50.120 and training 

elements from the annual training 

requirements in § 26.29(c)(2). The term 

‘‘systems approach to training’’ would 

have the meaning in proposed 

§ 53.725(c). A trainee assessment would 

be the same as in currently required 

SAT-based training programs. 

Proposed § 26.608(c) would require 

licensees and other entities to 

periodically evaluate their FFD training 

programs and revise them as 

appropriate. This training focus is not 

required by subpart K of part 26 or 

§ 26.29 but is proposed to address the 

flexibilities afforded in subpart M of 


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part 26. This section would be 

equivalent to § 50.120(b)(3). 

Proposed § 26.609 would require the 

implementation of a BOP. The proposed 

requirement would be equivalent to that 

in §§ 26.33 and 26.407, ‘‘Behavioral 

observation,’’ and would apply during 

construction, operation, and 

decommissioning, if applicable. Because 

subpart M of part 26 would apply 

during decommissioning through a 

licensee’s IMP, proposed § 26.609(a) and 

(b) were developed, in part, from 

proposed § 73.100(b)(9) and current 

§§ 73.55(b)(9) and 73.56(f) to help 

ensure consistency in the conduct of 

behavioral observation whether 

conducted for FFD or security purposes. 

Under the FFD program, the purpose 

of the BOP would be to help ensure that 

individuals subject to the FFD program 

are fit for duty and trustworthy and 

reliable to perform or direct those duties 

and responsibilities and maintain those 

types of access that make the individual 

subject to the FFD program. This 

assurance is accomplished by requiring 

each individual subject to subpart M of 

part 26 to be subject to behavioral 

observation, and by requiring all 

individuals to perform behavioral 

observation of others and report FFD 

concerns to the licensee- or other entitydesignated representative(s). The intent 

of the BOP requirement is not to require 

that all individuals be observed at all 

times by others; NRC-licensed operators, 

maintenance professionals, security 

officers, and others routinely perform 

solo operations periodically throughout 

the day. However, individuals must be 

subject to observation while they are 

performing or directing the performance 

of duties and responsibilities or 

maintaining the types of access making 

them subject to the FFD program. 

Observing behavior only at the 

beginning of a work shift is not 

sufficient to ascertain whether an 

individual is fit for duty, trustworthy, 

and reliable. Controlled substances may 

have a delayed effect between use (e.g., 

ingestion) and the onset of physiological 

or psychological effects, and fatigue 

accumulates with time. Behavior must 

be continually observed throughout the 

work shift to detect any changes from 

baseline human performance 

characteristics, including mental or 

physical health and mannerisms, or any 

activities that may indicate that the 

individual is not trustworthy and 

reliable. 

Proposed § 26.609(a) would differ 

from §§ 26.33 and 26.407 in that it 

would place the responsibility for 

performing behavioral observation on 

‘‘all individuals subject to this subpart,’’ 

rather than only those ‘‘individuals 

specified in § 26.4(f) [who] are 

constructing or directing the 

construction of safety- or securityrelated SSCs’’ in § 26.407 or 

‘‘individuals who are trained under 

§ 26.29 to detect behaviors’’ in § 26.33 to 

improve clarity. 

Proposed § 26.609(b) would require 

all individuals subject to the FFD 

program to report to the licensee- or 

other entity-designated representative 

any onsite or offsite behaviors or 

activities by individuals subject to this 

part that may constitute an 

unreasonable risk to the safety or 

security of the NRC-licensed facility or 

SNM or may cause harm to others. The 

NRC proposes this description of 

reportable conduct because an 

individual’s activities (e.g., use of illegal 

substances) and communications (e.g., 

hate speech or threats of violence) 

offsite are a direct indication of the 

individual’s fitness, trustworthiness, 

and reliability and must be evaluated as 

to whether authorization should be 

granted or maintained. Proposed 

§ 26.609(b) would include a description 

of this conduct instead of the § 26.33 

undefined phrase, ‘‘FFD concerns,’’ to 

enhance the clarity of the requirement. 

This proposed BOP reporting 

requirement would include any 

information relating to character or 

reputation of the individual indicating 

that the individual cannot be trusted or 

relied upon to perform those duties and 

responsibilities or maintain access to 

NRC-licensed facilities, SNM, or 

sensitive information. This would better 

align with the proposed § 73.120 BOP 

requirement, which states that each 

person subject to behavioral observation 

must communicate to the licensee or 

applicant observed behaviors or 

activities of individuals that may 

constitute an unreasonable risk to the 

health and safety of the public and 

common defense and security. Proposed 

§ 26.609(a) and (b) were written broadly 

to include offsite conduct that the 

reporting individual considers serious 

enough to call into question the 

character or reputation of the subject 

individual. 

Proposed § 26.609(c) would require 

that licensees and other entities perform 

behavioral observation visually, inperson, and, when necessary, remotely 

by live video and audible streaming and 

capture. This requirement was 

developed from the security observation 

requirements in § 73.55(e)(7)(i)(B) and 

(C), (h)(2)(v), and (i)(2) and (i)(5)(ii). 

Conducting an in-person observation of 

another individual is the preferred 

method to ascertain whether the 

observed individual can safely and 

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 § 26.609(d) 

regarding NRC-licensed operator 

manipulation of reactor controls. 

Additionally, certain duties (such as 

maintenance activities performed by a 

single worker outside of a control room) 

may not present an opportunity for 

video monitoring; in these situations, 

behavioral observation should be 

conducted on a sampling basis (i.e., a 

planned observation of the work 

activity) as outlined in a licensee’s or 

other entity’s FFD program. 

In situations involving small staff 

sizes, facilities sited in geographically 

remote locations, or both, additional 

observers would enhance the 

effectiveness of a BOP. Technological 

developments in automated safety and 

security systems may enable licensees 

or other entities to reduce staff sizes to 

10 to 40 percent of the staff size of an 

LWR facility licensed under part 50 or 

52. Smaller staff sizes may translate into 

more solo operations, less teamwork, 

fewer peer checks, or infrequent 

management oversight of field activities, 

leading to fewer behavioral 

observations. Therefore, a licensee or 

other entity would have fewer 

opportunities to observe whether 

individuals are fit for duty. Enabling 

video and audible streaming and 

capture to enhance the BOP would be 

consistent with the security-related 

behavioral observation requirement in 

proposed § 73.120(c)(2)(ii), which 

would also enable video conferencing or 

other acceptable electronic means 

promoting face-to-face interaction for 

those individuals working remotely. 

Proposed § 26.609(d) would require 

that licensees or other entities perform 

behavioral observation of NRC-licensed 

operators who manipulate the controls 

of any commercial nuclear plant 

licensed under part 53, remotely by live 

video and audible streaming capture for 

those part 53 facilities where individual 

task loading does not allow for the 

effective conduct of behavior 

observation in addition to assigned 

operational tasks. The purpose of this 

paragraph would be similar to that of 

proposed § 26.609(c), where the 

possibility of in-person observation is 

significantly diminished because of solo 

operations or because the facility may 

only require a minimum staff size 

onsite. 

Proposed § 26.610 would be 

equivalent to § 26.409, ‘‘Sanctions,’’ and 

would require the licensee or other 

entity to establish sanctions for FFD 


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policy violations that, at a minimum, 

prohibit the individuals specified in 

§ 26.4 from being assigned to perform or 

direct those duties and responsibilities 

or maintaining authorization making 

them subject to subpart M of part 26. To 

be consistent with § 26.75, ‘‘Sanctions,’’ 

the severity of the sanction as described 

in § 26.610 would escalate with the 

number of occurrences and severity of 

the FFD policy violation. The sanction 

would be long enough to help deter 

future FFD policy violations and 

facilitate counseling and treatment 

before the licensee reinstates the 

individual’s access to the facility. The 

NRC proposes this requirement because 

the 14-day denial described in § 26.75 

may not allow sufficient time for 

counseling and treatment based on the 

particular FFD policy violation. 

Equivalent to § 26.75(c), proposed 

§ 26.610 would also require a minimum 

5-year denial of access to the NRClicensed facility for certain violations of 

the FFD policy within the protected area 

of a commercial nuclear plant and by an 

individual or individuals who are the 

operators of the conveyance to transport 

or use formula quantities of strategic 

SNM. Equivalent to § 26.75(b), proposed 

§ 26.610 would require a permanent 

denial of authorization be issued for any 

subversion attempt. 

Proposed § 26.611 would protect 

information collected from FFD program 

implementation and would be 

equivalent to current § 26.411, 

‘‘Protection of information.’’ The 

protected information would include, 

but not be limited to, privacy and 

medical information. Section 26.611 

would not include the § 26.411 

requirement that FFD programs must 

maintain and use the personal 

information with the highest regard for 

individual privacy because such a 

requirement would be unnecessary in 

light of the proposed § 26.611(a) 

requirement that licensees and other 

entities must establish and maintain a 

system of files and procedures to 

prevent unauthorized disclosure. 

Proposed § 26.611(b), although 

equivalent to § 26.411(b), would require 

licensees and other entities to have all 

individuals sign a consent to be subject 

to the FFD program before subjecting 

the individual to the FFD program (e.g., 

before being subject to a pre-access test 

in § 26.607(b)(1), unlike § 26.411(b)). 

The purpose of this proposal would be 

to enhance protections afforded to 

individuals subject to the FFD program 

and their knowledge of, in part, why 

they are subject to drug and alcohol 

testing, behavioral observation, 

information collection, MRO reviews, 

and other FFD program elements. Like 

the consent required by § 26.411(b), the 

consent would authorize disclosure of 

the collected information. Consent 

would not be needed for disclosures to 

the individuals and entities specified in 

§ 26.37(b)(1) through (b)(6), (b)(8), and 

persons deciding matters under review 

in proposed § 26.613, ‘‘Appeals 

process.’’ 

Proposed § 26.613 would be 

equivalent to § 26.413, ‘‘Review 

process.’’ The proposed title was 

changed to an appeal process to clarify 

that § 26.613 would be the process 

implemented when an individual elects 

to appeal a licensee or other entity 

determination that the individual had 

violated the FFD policy. The proposal 

would also require that the process 

include a schedule for the completion of 

the review of the determination that the 

individual had violated the FFD policy. 

The NRC proposes this requirement 

because operating experience 

demonstrates that workers may not be 

protected from a continuous review 

process that does not result in an 

outcome. 

Proposed § 26.615 would require 

licensees and other entities to perform 

audits of the FFD program. The 

proposed section would be equivalent to 

§ 26.415, ‘‘Audits.’’ Under proposed 

§ 26.615(a), audits would be performed 

at a frequency that ensures the FFD 

program’s continuing effectiveness. This 

would be particularly important for FFD 

program elements that are not part of 

the FFD PMRP required by § 26.603(d). 

Corrective actions would be taken as 

soon as reasonably practicable to resolve 

any problems identified and preclude 

recurrence. Proposed § 26.615(b) would 

require the subject matter, scope, and 

frequency of audits be revised as 

necessary to improve or maintain 

program performance based on findings 

resulting from licensee or other entity 

implementation of its FFD PMRP. These 

requirements were developed from 

appendix B to part 50, ‘‘Quality 

Assurance Criteria for Nuclear Power 

Plants and Fuel Reprocessing Plants’’; 

criterion X, ‘‘Inspection’’; and criterion 

XVIII, ‘‘Audits.’’ 

Proposed § 26.615(c) would be 

equivalent to § 26.415(b) and would 

enable licensees and other entities to 

conduct joint audits or accept audits of 

C/Vs so long as the audit addresses the 

relevant services of the C/Vs. 

Proposed § 26.615(d) would be 

equivalent to § 26.415(c) by establishing 

requirements for the auditing of HHScertified laboratories. Unlike § 26.415(c), 

the proposal would not contain a 

reference to the Department of 

Transportation drug and alcohol testing 

requirements. This would broaden the 

regulatory flexibility afforded to a 

licensee or other entity in that they may 

use an offsite collection or testing 

facility that does not meet the 

Department of Transportation 

requirements. 

Proposed § 26.615(d) would state that 

licensees and other entities need not 

audit an HHS-certified laboratory if the 

licensee’s or other entity’s panel of 

drugs and drug metabolites to be tested 

is equivalent to the panel by which the 

laboratory is certified by HHS or is 

subject to the standards and procedures 

for drug testing and evaluation used by 

the laboratory under the HHS 

Guidelines. The NRC would afford this 

flexibility because the NRC is aware that 

HHS desires to streamline changes in its 

guidelines to its panel of drugs and drug 

metabolites to be tested. Therefore, if a 

licensee or other entity elects to 

implement the HHS Guidelines in its 

procedures and maintains the minimum 

panel of drugs and drug metabolites to 

be tested as required by subpart M of 

part 26, a licensee or other entity may 

still use (and not audit) the HHScertified laboratory because the 

§ 26.603(e) change control process 

would maintain FFD program 

effectiveness. 

To help ensure FFD program 

effectiveness, § 26.615(d) would also 

require that collection facility 

procedures are comparable to those 

required in subpart E of part 26, 

including a proposed requirement that 

the offsite facility’s specimen collection 

and testing procedures are audited on a 

biennial basis, which is also a 

protection consideration afforded to 

individuals subject to the FFD program. 

Conducting this audit on a biennial 

basis would be equivalent to that 

required in § 26.41(b) and would help 

ensure that the specimen collection 

process at the facility remains effective. 

Proposed § 26.617 would establish 

recordkeeping and reporting 

requirements equivalent to those in 

current § 26.417. However, § 26.617 

would require retention of records 

pertaining to administration of the FFD 

program and FFD performance data 

required by § 26.717 until license 

termination, which is based on current 

§ 26.711(a) because § 26.417 does not 

provide for a retention period. 

Proposed § 26.617(b)(1) would be 

identical to the reporting requirements 

in § 26.417(b)(1) regarding the licensee’s 

or other entity’s FFD program. 

Proposed § 26.617(b)(2) would require 

the reporting of annual (i.e., January 

through December) program 

performance information to the NRC 

before March 1 of the following year. 

This reporting would be equivalent to 


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13PNNL, Technical Letter Report, ‘‘The Use of 

Electronic Communications to Perform 

Determinations of Fitness,’’ dated August 2017. 

the annual program performance 

requirement in § 26.417(b)(1), and the 

March 1 due date is based on the 

reporting deadline in § 26.717(e). 

Licensees and other entities would be 

required to report FFD performance 

information using new NRC Forms 893, 

‘‘Single FFD Policy Violation Form,’’ 

and 894, ‘‘10 CFR part 26, subpart M, 

Annual Reporting Form for FFD 

Performance Information.’’ 

Proposed § 26.617(c) would require 

that FFD-related information be shared 

within the commercial nuclear industry 

when requested to support 

authorization determinations. This 

requirement would help individuals 

seeking employment by another NRClicensed facility subject to subpart C of 

part 26, complete their NRC-required 

sanctions and licensee-administered or 

-directed drug and/or alcohol abuse 

treatment plans before the restoration of 

authorization by a licensee or other 

entity. Information sharing may also 

enhance FFD program effectiveness 

because FFD-related lessons learned 

from, for example, substance testing, 

subversion attempts, and laboratory and 

MRO performance must be shared when 

requested. 

Proposed § 26.619 would require 

licensees or other entities to establish a 

process to evaluate individuals when 

their fitness or trustworthiness and 

reliability are in question. Section 

26.619 would be equivalent to § 26.419, 

‘‘Suitability and fitness 

determinations,’’ but, unlike § 26.419, 

would apply during the construction 

and operation phases. Also, proposed 

§ 26.619 would require that a suitability 

or fitness determination conducted for 

cause be conducted face-to-face. This 

proposed requirement is based on 

current § 26.189(c); however, unlike 

§ 26.189(c), proposed § 26.619 would 

not prohibit augmenting determinations 

via electronic means of communication. 

Instead, § 26.619 would explicitly 

permit determinations to be performed 

via electronic means, so long as those 

determinations are supported by an 

appropriately trained individual who is 

present in-person with the individual 

being assessed. 

In considering the current restriction 

on the use of electronic means of 

communication for determinations of 

fitness conducted for cause, the NRC 

finds that since publication of the 2008 

part 26 final rule, there have been 

developments in using electronic means 

of communication (i.e., 

‘‘videoconferencing’’) as an alternative 

to conducting face-to-face interactions. 

To address these considerations, the 

NRC contracted the Pacific Northwest 

National Laboratory (PNNL), DOE, to 

study whether a medical and mental 

health assessment via electronic 

communication could be an acceptable 

alternative to an in-person, face-to-face 

assessment.13 Based on this study, if 

electronic means were to be used to 

conduct a face-to-face assessment, an inperson element would still be integral to 

the assessment process. However, under 

certain circumstances, face-to-face 

determinations and assessments 

conducted as part of an FFD program for 

an entity licensed under part 53 (i.e., 

those determinations and assessments 

performed in accordance with § 26.619, 

§ 26.207, or § 26.211) may be augmented 

via electronic communications. Such 

remotely conducted determinations and 

assessments would be required to be 

conducted with someone who is present 

in-person with the individual being 

assessed and who is trained in 

accordance with the requirements of 

either § 26.29 and § 26.203(c) or § 26.608 

and § 26.202(c). Permitting the use of 

electronic communications would help 

ensure FFD program effectiveness, 

especially in instances where the part 

53 commercial nuclear plant is sited in 

a geographically remote location or 

when the facility has a small staff size. 

D. Proposed Changes to Part 26, Subpart 

N 

Proposed § 26.709 would make the 

recordkeeping and reporting 

requirements in subpart N of part 26 

applicable to licensees and other 

entities of facilities licensed under part 

53 that elect not to implement the 

requirements in subpart M of part 26 or 

elect to implement the requirements in 

§ 26.605(b). 

Proposed § 26.711(c) and (d) would be 

amended to make these requirements 

applicable to licensees or other entities 

described in § 26.3(f). Section 26.711(c) 

provides protection to individuals 

subject to part 26 by enabling an 

individual’s right to review FFD-related 

information and correct any inaccurate 

or incomplete information. Section 

26.711(d) requires, in part, that any 

FFD-related information shared with 

other licensees or other entities is 

correct and complete. 

E. Proposed Changes to Part 26, Subpart 

O 

The vast majority of the proposed 

changes to part 26 would be new or 

revised substantive provisions that 

would establish a regulatory obligation 

or prohibition or would be conforming 

edits to reflect the addition of part 53. 

The only new provision that would not 

be substantive, such that violation of it 

would not result in a criminal penalty, 

would be proposed § 26.601. Therefore, 

the NRC proposes to add § 26.601 to the 

list of regulations in § 26.825(b) to 

which criminal sanctions do not apply. 

10 CFR Part 50 

A. Section 50.160: Emergency 

Preparedness for Small Modular 

Reactors, Non-Light-Water Reactors, 

and Non-Power Production or 

Utilization Facilities 

This proposed rule would revise 

§ 50.160(b)(3) and (c)(2) to make that 

section applicable to applicants and 

licensees under part 53. Section 50.160 

provides an alternative to other part 50 

emergency preparedness requirements 

focused on large light-water reactors to 

provide an optional emergency 

preparedness framework specifically for 

small modular reactors (SMRs) and 

other new technologies. These 

alternative emergency preparedness 

requirements adopt a performancebased, technology-inclusive, riskinformed, and consequence-oriented 

approach. Commercial nuclear reactor 

applicants complying with § 50.160 

would be required to submit as part of 

the application the analysis used to 

determine whether the criteria in 

§ 53.1109(g)(2)(i)(A) and (B) are met 

and, if they are met, the size of the 

plume exposure pathway emergency 

planning zone (EPZ). An EPZ bounds 

the area surrounding a facility within 

which detailed planning is needed to 

implement predetermined, prompt 

protective actions. The criterion in 

proposed § 53.1109(g)(2)(i)(A) is that 

public dose, as defined in § 20.1003, is 

projected to exceed 10 mSv (1 rem) 

TEDE over 96 hours from the release of 

radioactive materials from the facility 

considering accident likelihood and 

source term, timing of the accident 

sequence, and meteorology. The 

criterion in proposed 

§ 53.1109(g)(2)(i)(B) is that predetermined, prompt protective measures 

are necessary. These are the same 

criteria that are in § 50.33(g)(2)(i)(A) and 

(B) and are used to assess the need for 

and size of an EPZ in applications under 

parts 50 and 52. 

Applicants choosing to comply with 

§ 50.160 must determine the 

radiological releases from the facility 

that are evaluated in the determination 

of the plume exposure pathway EPZ. 

Consistent with other Federal guidelines 

such as the Federal Emergency 

Management Agency ‘‘Radiological 

Emergency Preparedness Program 

Manual,’’ issued in 2023, and the 


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Environmental Protection Agency ‘‘PAG 

Manual: Protective Action Guides and 

Planning Guidance for Radiological 

Incidents,’’ issued in 2017, applicants 

should consider quantitative and 

qualitative information on the potential 

radiological releases that make up the 

spectrum of accidents used to develop 

the basis for the applicant’s site-specific 

EPZ. This information is derived from 

the licensing basis. The NRC plans to 

update the risk-informed approach in 

RG 1.242 for part 53 while maintaining 

its flexibility for using information 

already developed and available in 

licensing basis documents, including 

PRA results, deterministic dose 

quantities, accident timing, target set 

analyses, mitigation capabilities, and 

site-specific factors such as 

meteorology. 

In its safety analysis report, the 

applicant would describe the LBEs 

relevant to the facility and would 

consider these LBEs as candidates for 

the spectrum of accidents used to 

develop the site-specific EPZ. The LBEs 

assessed include a wide range of events 

that are appropriate for considering in 

the facility’s emergency preparedness 

and response planning. In addition, 

§ 50.160(b)(1)(iv)(A)(2) requires 

licensees to be capable of implementing 

their approved emergency response plan 

in conjunction with their safeguards 

contingency plan. Radiological sabotage 

events are typically factored into EPZ 

determinations by considering 

consequences to be bounded by LBEs 

and by crediting protection against the 

DBT in reducing the likelihood of a 

release. 

The provisions in proposed 

§ 53.860(a) provide an alternative to 

applicants and licensees by not 

requiring them to protect against the 

DBT of radiological sabotage in 

accordance with §§ 73.55 and 73.100 if 

they can demonstrate that the 

consequences from unmitigated 

radiological sabotage events are below 

the safety criteria in proposed § 53.210. 

The deployment of some commercial 

nuclear plants under part 53 may 

involve new scenarios where the source 

terms and consequences of sabotagerelated events are not bounded by the 

consequences of the unlikely and very 

unlikely event sequences analyzed 

under subpart C. Accordingly, the NRC 

plans to develop guidance for part 53 

applicants and licensees choosing to 

comply with the alternative emergency 

preparedness requirements in § 50.160 

to address this new class of reactors. In 

Section VI of this document, the NRC is 

asking for stakeholder feedback on the 

clarity of the regulations and guidance 

for various scenarios that might arise in 

implementing graded approaches for 

security and emergency planning for 

some commercial nuclear plant designs. 

B. Appendix B to Part 50: Quality 

Assurance Criteria for Nuclear Power 

Plants and Fuel Reprocessing Plants 

Appendix B to part 50 would be 

amended to make it applicable to 

applicants and licensees under part 53. 

This results in the need for some 

revisions to recognize differences in 

terminology between parts 50 and 53. 

Namely, the term ‘‘design bases,’’ which 

is defined in § 50.2, is not used in part 

53. For this reason, text is added in both 

Section III, ‘‘Design Control,’’ and 

Section IV, ‘‘Procurement Document 

Control,’’ to refer to ‘‘functional design 

criteria, as defined in § 53.020,’’ as the 

part 53 equivalent of the term ‘‘design 

bases.’’ 

10 CFR Part 73 

A. Section 73.100: Technology-Inclusive 

Requirements for Physical Protection of 

Licensed Activities at Commercial 

Nuclear Plants Against Radiological 

Sabotage 

Proposed § 73.100 would provide a 

performance-based regulatory 

framework for the design, 

implementation, and maintenance of a 

physical protection program and 

security organization for certain 

commercial nuclear plants licensed 

under part 53. The current § 73.55 

physical security requirements for 

nuclear power reactors licensed under 

part 50 and part 52 use a combination 

of performance criteria (e.g., 

§ 73.55(b)(1) through (3)) and numerous 

prescriptive requirements developed to 

achieve performance objectives (e.g., 

§ 73.55(k)(5)(ii)). By contrast, in the 

proposed performance-based approach 

to physical security for part 53, 

performance objectives and 

requirements would be the primary 

bases for regulatory decision-making, 

giving the licensee the flexibility to 

determine how to demonstrate 

compliance with the established 

performance criteria for an effective 

physical protection program. This 

proposed physical protection program 

would provide an optional pathway for 

licensees that elect not to demonstrate 

compliance with the provisions in 

§ 73.55 and do not satisfy the criterion 

as described in proposed § 53.860(a)(2). 

This proposed physical protection 

program would provide that activities 

involving SNM are not inimical to the 

common defense and security and do 

not constitute an unreasonable risk to 

the public health and safety. 

Section 73.100(a) would require each 

part 53 licensee that elects to 

demonstrate compliance with this 

section rather than § 73.55 to implement 

the requirements therein through its 

physical security plan, training and 

qualification plan, safeguards 

contingency plan, and cybersecurity 

plan (referred to collectively hereafter as 

‘‘security plans’’) prior to initial fuel 

load into the reactor (or, for a fueled 

manufactured reactor, before initiating 

the physical removal of any one of the 

independent physical mechanisms to 

prevent criticality required under 

§ 53.620(d)(1)). The security plans 

would need to identify, describe, and 

account for site-specific conditions that 

affect the licensee’s capability to satisfy 

the requirements of § 73.100. Based on 

experience from recent new reactor 

licensing reviews, the NRC recognizes 

that licensees may seek to receive 

unirradiated fuel onsite before carrying 

out the security requirements in 

§ 73.100. However, these security 

requirements would have to be 

implemented at some point before 

reactor operation to address the 

increased risk arising from irradiated 

fuel onsite. This proposed rule would 

make clear that part 53 applicants and 

licensees using § 73.100 may bring 

unirradiated nuclear fuel onsite and 

protect it in accordance with the NRC’s 

requirements for physical protection of 

SNM of moderate and low strategic 

significance under § 73.67 until initial 

fuel load into the reactor (or, for a fueled 

manufactured reactor, until initiating 

the physical removal of any one of the 

independent physical mechanisms to 

prevent criticality required under 

§ 53.620(d)(1)). 

Section 73.100(b) would outline the 

general performance objective and 

design requirements of the licensee 

physical protection program. A 

licensee’s program would be required to 

provide protection against any 

deliberate act within the DBT of 

radiological sabotage, including spent 

fuel sabotage, which could directly or 

indirectly endanger the public health 

and safety by exposure to radiation. The 

physical protection program is 

supported by the AA program, 

cybersecurity program, and IMP to 

demonstrate compliance with the 

general performance objective of 

§ 73.100(b). 

Section 73.100(b)(2) was developed, 

in part, from § 73.55(b)(3). To satisfy the 

general performance objective of 

§ 73.100(b)(1), the physical protection 

program would need to protect against 

the DBT of radiological sabotage. The 

existing fleet of LWR satisfies this 

objective by preventing significant core 


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14The 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, TechnologyInclusive Regulatory Framework for Advanced 

Reactors (RIN 3150–AK31),’’ dated March 4, 2024, 

the Commission approved a similar approach to 

defend against radiological sabotage. 

damage and spent fuel sabotage. Some 

non-LWR reactor licensees’ physical 

protection programs may be designed to 

prevent a significant release of 

radionuclides from any source. 

Therefore, the proposed performance 

objective would focus on radiological 

sabotage in general, rather than a 

specific focus on core damage or spent 

fuel sabotage, to be technology inclusive 

and allow for flexibility for different 

reactor technologies. 

Under the proposed § 73.100(b)(2)(ii), 

licensees must provide defense in depth 

in achieving performance requirements 

through the integration of engineered 

systems, administrative controls, and 

management measures. This 

requirement would apply defense-indepth concepts as part of the physical 

protection program to ensure the 

capability to demonstrate compliance 

with the performance objective of the 

proposed § 73.100(b)(1) is maintained in 

the changing threat environment. The 

defense-in-depth philosophy applies to 

measures against intentional acts as 

required by § 73.100(b), and the designs 

of physical security systems should 

employ defense in depth through 

systems diversity, independence, and 

separation under § 73.100(b)(2). The 

most common defense-in-depth 

measures apply concepts of 

redundancy, diversity, independence, 

and safety margin to ensure systems 

reliability and availability. The defensein-depth philosophy applies to the 

design of a physical protection program, 

which integrates engineered controls 

and administrative controls, to provide 

protection against the DBT for 

radiological sabotage. 

Section 73.100(b)(3) would require 

the physical protection program to be 

designed and implemented to achieve 

and maintain the reliability and 

availability of SSCs required for 

demonstrating compliance with 

specified performance requirements. 

These physical protection performance 

requirements were informed by 

§ 73.55(b) and the Commission’s 

Advanced Reactor Policy Statement. 

The performance objective of 

protecting against the DBT of 

radiological sabotage is achieved by the 

design and implementation of the 

physical protection program, 

maintained at all times, with the 

following required performance 

capabilities proposed in the provisions 

in § 73.100(b)(3): intrusion detection, 

intrusion assessment, security 

communication, security response, 

protecting against land and waterborne 

vehicle bomb assaults, and access 

control portals. The physical protection 

program must maintain the reliability 

and availability of SSCs relied upon for 

demonstrating compliance with the 

performance requirements. The terms 

‘‘reliability and availability’’ are 

intended to describe defense in depth in 

a performance-based manner and would 

be critical elements for demonstrating 

compliance with the proposed 

requirement for protection against the 

DBT of radiological sabotage as 

described in the proposed 

§ 73.100(b)(2). 

The first element, ‘‘intrusion 

detection,’’ would be provided through 

the use of detection equipment, patrols, 

access controls, and other program 

elements and would provide 

notification to the licensee that a 

potential threat is present and where the 

threat is located. 

The second element, ‘‘intrusion 

assessment,’’ would provide a 

mechanism through which the licensee 

would identify the nature of the threat 

detected. This would be accomplished 

through the use of video equipment, 

patrols, and other program elements that 

would provide the licensee with timely 

information about the threat for use in 

determining how to respond. 

The third element, ‘‘security 

communication,’’ would provide a 

mechanism through which the licensee 

would communicate the necessary 

information to the response force to 

ensure effectiveness of the physical 

protection program. This would be 

accomplished through the redundant, 

independent, and diverse design of 

physical security and/or plant SSCs 

relied on for onsite and offsite security 

communications. The continuity and 

integrity of communications should 

account for the DBT’s ability to affect 

the reliability and availability of 

communications. 

The fourth element, ‘‘security 

response,’’ would provide a mechanism 

through which the licensee would be 

capable of timely security response to 

interdict and neutralize threats up to 

and including the DBT of radiological 

sabotage. The security response may 

include the use of onsite armed 

responders, law enforcement responders 

(local, State, or Federal), or other offsite 

armed responders (e.g., licensee 

proprietary or contract security 

personnel who are positioned offsite), or 

a combination thereof, as appropriate.14

The licensee must provide protection 

against any element of the DBT, to 

include those that do not rise to the full 

capability of the DBT. Structures, 

systems, and components relied on to 

provide delay functions must be 

designed to provide for timely response 

to adversary attacks with adequate 

defense in depth. Delay would allow the 

licensee to take necessary actions to 

counter any attempt by the threat to 

advance towards the protected target or 

target set element. The overall response 

objective would be to place the threat in 

a condition from which the threat no 

longer has the potential for, or 

capability of, doing harm to the 

protected target. 

The fifth element, ‘‘protecting against 

land and waterborne vehicle bomb 

assaults,’’ would provide a mechanism 

through which the licensee would be 

capable of protecting the plant against 

the DBT vehicle bomb assault. The 

methods that are relied on to protect 

against a DBT land vehicle and 

waterborne vehicle bomb assault must 

be designed to protect the reactor 

building, structures containing safety or 

security related systems, and 

components from explosive effects. 

The sixth element, ‘‘access control 

portals,’’ would provide a mechanism 

through which the licensee would be 

capable of detecting and denying 

unauthorized access to persons and 

pass-through of contraband materials 

(e.g., weapons, incendiaries, explosives) 

to protected areas. Integrity of the access 

control system is maintained through 

licensee oversight and ensures that 

attempts to circumvent or bypass the 

established process will be detected and 

access denied. 

The proposed performance 

requirements would permit the 

applicant or licensee to determine how 

to design the physical protection 

program to protect the plant against the 

DBT of radiological sabotage without 


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prescriptive requirements such as those 

currently found in § 73.55. DG–5076, 

‘‘Guidance for Technology Inclusive 

Requirements for Physical Protection of 

Licensed Activities at Commercial 

Nuclear Plants,’’ has been developed by 

the NRC to describe one acceptable 

approach to demonstrate compliance 

with requirements proposed in § 73.100. 

Section 73.100(b)(4) would require 

the licensee to identify target sets in 

accordance with § 73.55(f). For nonLWR and SMRs, target sets would be 

defined in DG–5071, ‘‘Target Set 

Identification and Development for 

Nuclear Power Plants,’’ as the minimum 

combination of equipment, operator 

actions, and/or structures that, if all are 

prevented from performing their 

intended safety function or prevented 

from being accomplished, barring 

extraordinary actions by plant 

operations, would likely result in a 

significant release of radionuclides from 

any source (e.g., a release to the 

environment exceeding that analyzed in 

the DBA licensing basis). 

Section 73.100(b)(5) would require 

that each licensee perform a site-specific 

analysis for the purpose of identifying 

and analyzing site-specific conditions 

that affect the design of the onsite 

physical protection program. 

Section 73.100(b)(6) would require 

licensees to implement a performance 

evaluation program, which would 

ensure that a licensee will periodically 

test and evaluate the effectiveness of the 

physical protection program to protect 

against the DBT. This program would 

ensure that licensees are able to 

demonstrate that the physical protection 

program satisfies the response 

requirements of § 73.100 and that the 

site’s protective strategy effectively 

protects against the DBT. Licensee 

performance evaluations would include 

methods to assess, test, and challenge 

the integration of the physical 

protection programs functions and 

demonstrate the effectiveness of security 

plans, licensee protective strategy, and 

implementing procedures in accordance 

with § 73.100(g). 

Section 73.100(b)(7) would require 

licensees to implement an AA program 

in accordance with § 73.56. Section 

73.100(b)(8) would require licensees to 

establish, maintain, and implement 

protection against a cyberattack based 

on either the proposed cybersecurity 

program described in § 73.110 or the 

program described in existing § 73.54. 

Section 73.100(b)(9) would require an 

IMP that monitors the initial and 

continuing trustworthiness and 

reliability of individuals granted or 

retaining unescorted access or 

unescorted AA to a protected or vital 

area. The IMP must also implement 

defense-in-depth methodologies to 

minimize the potential for an insider 

(active, passive, or both) to adversely 

affect the licensee’s capability to protect 

against radiological sabotage. Because 

no one element of the AA program, FFD 

program, cybersecurity program, or 

physical protection program, would, by 

itself, provide the level of protection 

against the insider necessary to 

demonstrate compliance with the 

performance objective of the proposed 

§ 73.100(b), the effective integration of 

these programs is a necessary 

requirement to achieve defense in depth 

against the potential insider. 

Section 73.100(b)(10) would require 

that the licensee have the capability to 

track, trend, correct, and prevent 

recurrence of failures and deficiencies 

in the implementation of the 

requirements of this section. Section 

73.100(b)(11) would require the 

coordination of the security plans and 

associated procedures with other onsite 

plans to manage the safety and security 

interface during normal or emergency 

operations. 

Section 73.100(c) was developed from 

§ 73.55(c)(7), ‘‘Security implementing 

procedures,’’ and § 73.55(d), ‘‘Security 

organization,’’ and would outline the 

requirements for the composition, 

equipping, and training of the security 

organization. The purpose of the 

security organization is to effectively 

implement the physical protection 

program. Individuals assigned to 

perform physical protection or 

contingency response duties must be 

trained, equipped, and qualified to 

perform assigned duties and 

responsibilities. 

Section 73.100(d) would establish a 

performance requirement for searches of 

personnel, vehicles, and materials for 

the protection against radiological 

sabotage. The requirement describes 

broad categories of material (explosives, 

firearms, incendiary devices, etc.) to be 

detected and prevented from entry into 

the protected area; specific items that 

will be prohibited would not be 

prescribed in the regulation but will be 

stated in the licensee security plans 

with detailed descriptions being 

identified in implementation 

procedures. 

Section 73.100(e) would require a 

training and qualification program, 

described in the training and 

qualification plan, that ensures 

personnel are able to effectively perform 

their assigned security-related job 

duties. This high-level requirement 

would allow flexibility in how the 

licensee chooses to train its security 

personnel. One method for 

accomplishing this requirement would 

be to provide a training and 

qualification program that would be 

equivalent to appendix B to part 73. 

Section 73.100(f) would require 

periodic security reviews of the physical 

protection program to ensure effective 

implementation of the program by 

independent individuals. The 

evaluation process would provide a 

systematized approach for assessing the 

physical protection program as a basis 

for further development and 

improvement. Program reviews should 

be designed to ensure that the physical 

protection program maintains 

effectiveness and demonstrates 

compliance with NRC requirements. 

Section 73.100(f)(1) was developed from 

§ 73.55(m) and would require review of 

each element of the physical protection 

program. Section 73.100(f)(2) would 

require licensees to perform selfassessments of physical protection 

program functions to ensure that the 

capability to detect, assess, interdict, 

and neutralize the DBT of radiological 

sabotage is maintained. Section 

73.100(f)(3) would require an audit of 

the effectiveness of the physical 

protection program; security plans; 

implementing procedures; cybersecurity 

programs; management of the safety/ 

security interface activities; the testing, 

maintenance, and calibration program; 

and response commitments by local, 

State, and Federal law enforcement 

authorities. Section 73.100(f)(4) would 

require that results and 

recommendations, management 

findings, and any actions taken be 

documented and maintained to be 

available for inspection by the NRC. 

These reviews are independent of the 

ongoing performance evaluations 

described in § 73.100(b)(6) and (g). 

Section 73.100(g) would require that 

licensee performance evaluations, 

described in § 73.100(b)(6), include 

methods appropriate and necessary to 

assess, test, and challenge the 

integration of the physical protection 

program’s functions to protect against 

the DBT. The performance evaluations 

must also address the licensee’s 

measures to protect against cyberattacks, 

in accordance with the required 

cybersecurity plan, and engineered 

systems designed to protect against the 

DBT standalone ground vehicle bomb 

attack. 

Section 73.100(h) would establish 

performance requirements for 

maintaining security SSCs relied on to 

perform security functions to protect 

against the DBT. It would require that 

corrective actions and compensatory 

measures be taken by a licensee in 

response to a degradation of security 


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equipment or failure of the equipment 

to perform its intended functions. The 

licensee would be required to maintain 

the SSCs described in its design and 

licensing basis to ensure that they are 

reliable and available. 

Section 73.100(i) would establish 

requirements for the suspension of 

security measures in response to 

emergency and extraordinary 

conditions. The requirements of this 

paragraph, which were developed from 

§ 73.55(p), would be intended to 

provide flexibility to a licensee for 

taking reasonable actions that depart 

from a security plan in an emergency 

when such actions are immediately 

needed to protect the public health and 

safety and no action consistent with 

license conditions and TS that can 

provide adequate or equivalent 

protection is immediately apparent in 

accordance with proposed § 53.740(h). 

Section 73.100(j) would establish 

requirements regarding the inspection, 

retention and maintenance of records 

required to be kept by the NRC 

regulations, orders, or license 

conditions. These proposed 

requirements are developed from 

§ 73.55(q). 

B. Section 73.110: Technology-Inclusive 

Requirements for Protection of Digital 

Computer and Communication Systems 

and Networks 

Section 53.860 would require that a 

licensee establish, implement, and 

maintain a cybersecurity program in 

accordance with § 73.54 or § 73.110. 

Section 73.110 would establish 

requirements for the development and 

maintenance of a cybersecurity program 

for commercial nuclear plants licensed 

under part 53. This proposed section 

would implement a graded approach to 

determine the level of cybersecurity 

protection required for digital 

computers, communication systems, 

and networks. The proposed new 

section is informed by: (1) the operating 

experience from power reactors and fuel 

cycle facilities; and (2) the existing 

§ 73.54 framework, which addresses 

some of the basic issues for 

cybersecurity regardless of the type of 

reactor. Differences between the § 73.54 

requirements and those proposed in 

§ 73.110 are primarily based on the 

implementation of a consequence-based 

approach to cybersecurity that provides 

flexibility to accommodate the wide 

range of reactor technologies to be 

assessed by the NRC. A graded approach 

based on consequences is intended to 

account for the differing risk levels 

among reactor technologies. 

Specifically, the proposed new section 

would require licensees to demonstrate 

protection against cyberattacks in a 

manner that is commensurate with the 

potential consequences from those 

attacks. 

Under proposed § 73.110(a), licensees 

would need to ensure that digital 

computer and communications systems 

are adequately protected against a 

potential cyberattack that would result 

in: (1) a scenario where the cyberattack 

leads to offsite radiation doses that 

would endanger public health and 

safety (i.e., the resulting consequence 

exceeds the reference dose values in 

§ 53.210); or (2) a scenario where the 

cyberattack adversely impacts the 

physical security digital assets used by 

the licensee to prevent unauthorized 

removal of material or radiological 

sabotage. Security digital assets would 

include those used for nuclear MC&A. 

The proposed § 73.110(b) would 

require licensees to protect the 

communication system and networks 

associated with the functions described 

in § 73.110(a)(1) and (a)(2) from 

cyberattacks. To accomplish this, the 

licensee would establish, implement, 

and maintain a cybersecurity program 

for protecting digital assets within the 

scope of § 73.110 that would make use 

of risk insights, including threat 

information, and would consider the 

resulting level of consequences of the 

threats. If the outcome of the assessment 

by the licensee under § 73.110(b)(1) 

revealed that a potential cyberattack 

would not compromise any digital 

assets that support safety and security 

functions, and thus would not result in 

the consequences listed in § 73.110(a) 

(e.g., would not exceed the reference 

dose values), then only a narrow set of 

the cybersecurity program requirements 

in § 73.110(d) and (e) would apply. For 

example, the licensee would only need 

to develop a cybersecurity program that 

implements the requirements dealing 

with: 

• Analyzing modifications of any 

asset before implementation to see if 

they demonstrate compliance with the 

potential consequences in § 73.110(a); 

• Ensuring employees and contractors 

are aware of cybersecurity requirements 

and have some level of cybersecurity 

training; 

• Evaluating and managing 

cybersecurity risks to the plant; 

• Reviewing the cybersecurity plan 

for any required changes; and, 

• Retaining records of the 

cybersecurity plan along with any plan 

changes. 

Section 73.110(c) through (e) were 

developed from § 73.54(a)(2), and (c) 

through (h), respectively. 

The proposed requirements would 

address the need for the licensee to 

develop a cybersecurity program that 

implements a defense-in-depth 

protective strategy as required by 

proposed section § 73.110(d)(2). A 

defense-in-depth protective strategy for 

cybersecurity is represented by 

collections of complementary and 

redundant security controls that 

establish multiple layers of protection to 

safeguard critical digital assets. Under a 

defense-in-depth protective strategy, the 

failure of a single protective strategy or 

security control should not result in the 

compromise of safety and security 

functions. 

C. Section 73.120: Access Authorization 

Program for Commercial Nuclear Plants 

Section 73.120 would address AA for 

certain commercial nuclear plants 

licensed under part 53. The proposed 

language in § 73.120 would provide an 

alternate approach to the existing 

framework for AA under §§ 73.55, 

73.56, and 73.57, commensurate with 

risk and consequences to public health 

and safety. It would be available to part 

53 applicants and licensees who 

demonstrate in an analysis that the 

offsite consequences of a DBE satisfy the 

criterion defined in § 53.860(a)(2)(i) (i.e., 

would not exceed the offsite dose values 

in § 53.210(b)). The proposed 

requirements in § 73.120 would be 

similar to the existing AA program 

elements for those NRC licensed 

facilities issued additional security 

measures (ASMs) orders and for 

materials licensees under § 37.21. 

Applicants not satisfying the criterion 

would need to establish, implement, 

and maintain a full AA program, 

including an IMP, in accordance with 

§ 73.56. 

Proposed § 73.120(a) would be based 

on an applicant satisfying the eligibility 

criterion in § 53.860(a)(2)(i). Section 

73.120(b) would identify the categories 

of individuals who would be subject to 

an AA program in accordance with this 

section. The applicability statement in 

§ 73.120(b)(1)(i) would encompass 

individuals whom the licensee intends 

to grant unescorted access to the 

facilities’ most sensitive areas, 

consistent with § 73.56(b)(1)(i) for 

power reactors and the ASM orders and 

license conditions issued to any NRC 

licensed facility or material licensee. 

Sections 73.120(b)(1)(ii) through (iv) 

would be consistent with 

§ 73.56(b)(1)(ii) through (iv), 

respectively. The program would 

include individuals who may be onsite 

or offsite (e.g., remote operators or 

information technology staff) and have 

virtual access to important plant 

operational and communication systems 

based upon assigned duties and 


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responsibilities. An individual who has 

remote access to plant equipment and 

communication systems may have 

trusted privileges greater than the 

personnel at the plant site. Section 

73.120(b)(1)(iii) would state that offsite 

law enforcement personnel on official 

duty would not be subject to the 

licensee AA program. 

Section 73.120(c) would provide 

general performance objectives and 

requirements largely consistent with the 

AA program requirements for nuclear 

power reactors under § 73.56 and would 

provide licensees and applicants the 

flexibility in establishing their AA 

program to demonstrate compliance 

with various performance objectives. 

Section 73.120(c)(1) would include 

background investigation requirements 

consistent with § 37.25, as well as ASMs 

and license conditions that are applied 

to non-power reactor licensees. 

Background investigations include 

important elements to establish the 

trustworthiness and reliability of an 

individual, such that they do not 

constitute an unreasonable risk to 

public health and safety or the common 

defense and security. These include the 

following: (1) personal history 

disclosure, (2) verification of true 

identity, (3) employment history 

evaluation, (4) unemployment/military 

service/education, (5) credit history 

evaluation, (6) character and reputation 

evaluation, and (7) Federal Bureau of 

Investigation criminal history record 

check. 

Section § 73.120(c)(2) would establish 

behavioral observation requirements, 

which are an awareness initiative for 

recognizing behaviors adverse to the 

safe operation and security of the 

facility through observing the behavior 

of others in the workplace and reporting 

aberrant behavior or changes in 

behavior that might reflect negatively on 

an individual’s trustworthiness or 

reliability. Maintaining behavioral 

observation would assist and/or 

improve worker safety and reduce the 

risk of an insider threat. This proposed 

requirement in § 73.120(c)(2) would be 

a scaled version of the full BOP required 

under § 73.56(f). 

Section § 73.120(c)(2) would provide 

licensees greater flexibility to 

implement behavioral observation 

options for individuals granted 

unescorted access to the commercial 

nuclear plant’s protected area. Such 

options on reporting questionable 

behavior may include a program similar 

to the Department of Homeland 

Security’s program, ‘‘If you see 

something, say something,’’ or to a 

corporate behavioral awareness 

program. Commensurate with the 

potential lower safety and security risks 

of a commercial nuclear plant that 

meets the criterion in § 53.860(a)(2)(i), 

§ 73.120(c)(2) would not require the 

establishment of a comprehensive 

training program for behavioral 

observation (i.e., initial and refresher 

training including knowledge checks) as 

required for power reactors under 

§ 73.56 and part 26. Under 

§ 73.120(c)(2)(ii), behavioral observation 

would be able to be performed in-person 

or remotely by video, and identified 

behavior of concern would need to be 

reported to plant supervision. The 

remote access alternative to face-to-face 

interactions provides substantial 

flexibility for licensees and applicants. 

Any video conferencing or other 

acceptable electronic means promoting 

face-to-face interaction for those 

individuals working remotely would 

demonstrate compliance with this 

regulation. 

Section 73.120(c)(3) captures and 

maintains the self-reporting of legal 

actions as an essential performance 

element to enhance the licensee’s 

behavioral observation initiative similar 

to the current requirements under 

§ 73.56(g), assuring that personnel who 

are granted and who maintain 

unescorted access are trustworthy and 

reliable. 

Section 73.120(c)(4) would provide a 

scalable approach for granting and 

maintaining unescorted access. One 

component not included from § 73.56 is 

the need for a psychological assessment 

and reassessment under § 73.56(e) for 

granting unescorted access and 

§ 73.56(i)(v)(B) for individuals who 

perform one or more of the job functions 

described in § 73.120(b)(1)(ii) for 

maintaining unescorted access. 

Moreover, the requirement would 

permit criminal history updates to be 

completed within 10 years of the last 

review, compared to the three- or fiveyear reinvestigation periodicity for 

personnel at an operating commercial 

nuclear plant. In addition, no credit 

check re-evaluation would be required 

for these individuals. 

The continued need to maintain 

unescorted access would be evaluated 

on an annual basis by the reviewing 

official. Guidance in DG–5074, ‘‘Access 

Authorization Program for Commercial 

Nuclear Plants,’’ would specify that this 

evaluation should be based on a 

compilation of personnel interactions as 

described in the licensee’s or applicant’s 

policy and procedures for behavioral 

observation and the maintenance of an 

approved AA list. 

Section 73.120(c)(5) would require 

licensees and applicants to determine 

when a person no longer requires the 

need for unescorted access or no longer 

satisfies the AA requirement found 

within this section. Guidance in DG– 

5074 would further explain that 

licensees have the flexibility to 

terminate unescorted access to specific 

areas of the site if individuals lack the 

continued need for that access to 

perform their duties and 

responsibilities. 

Section 73.120(c)(6) would be 

consistent with the purpose of § 37.23(e) 

and would include the individual’s 

right to correct and complete 

information as required under 

§ 37.23(g). The section would include a 

requirement for designating a reviewing 

official. The language would provide 

clarity regarding the roles and 

responsibility of a reviewing official, 

who would be the only individual 

authorized to make unescorted access 

determinations. 

Section 73.120(c)(7) would align with 

the corresponding requirements under 

§ 37.23(f), and § 73.120(c)(8) would 

align with the corresponding 

requirements under § 37.31. These 

requirements would encompass the 

roles and responsibilities for licensees, 

applicants, and if applicable, the 

contractor/vendors to establish, 

implement, and maintain a system of 

files and records to ensure personal 

information is not disclosed to 

unauthorized persons. 

Section 73.120(c)(9) would align with 

the requirements of § 37.33. Section 

73.120(c)(10) would require licensees, 

applicants, and contractors or vendors 

to maintain the records that are required 

by the regulations in this section and 

retain them for a period of 3 years after 

the record is superseded or no longer 

needed. The record retention period of 

three years would be consistent with 

§ 37.23(h), contrasting with the five-year 

retention period under § 73.56(o). 

Records maintained in any database(s) 

would need to be available for NRC 

review, consistent with the 

requirements found under 

§ 73.56(o)(6)(ii). 

VI. Specific Requests for Comments 

The NRC is seeking advice and 

recommendations from the public on 

this proposed rule. We are particularly 

interested in comments and supporting 

rationale from the public on the 

following: 

Part 26—Fitness for Duty Program 

1. The proposed rule under 

§ 26.603(c) would enable a licensee or 

other entity to implement an FFD 

program under proposed § 26.604, ‘‘FFD 

program requirements for facilities that 

satisfy the § 26.603(c) criterion,’’ if the 


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licensee or other entity performs a sitespecific analysis to demonstrate that the 

facility and its operation satisfy the 

criterion in § 53.860(a)(2). 

Should the NRC consider replacing its 

proposed § 26.603(c) criterion 

referencing § 53.860(a)(2) with an 

alternative requirement that if the 

commercial nuclear plant is of the class 

described in § 53.800, ‘‘Facility 

licensees for self-reliant-mitigation 

facilities,’’ and either § 53.800(a)(1) or 

(2) is satisfied, then drug and alcohol 

testing would not be required? This 

proposal would align the § 26.603(c) 

criterion with that proposed in the NRClicensed operator regulatory framework 

of part 53. Please provide your 

considerations and rationale for your 

recommendation. 

Should the NRC also consider making 

a conforming change to the proposed 

§ 73.120 criterion used for the AA 

program? Please provide your 

considerations and rationale for your 

recommendation. 

Part 26—Technology-Inclusive 

Approaches to Fatigue Management 

Requirements Applicable to Unit 

Outages 

In establishing the outage minimum 

days off requirement of § 26.205(d)(4), 

the NRC’s objective was to ensure that 

individuals performing the duties 

described in § 26.4(a)(1) through (a)(4) 

have sufficient periodic long-duration 

breaks to prevent cumulative fatigue 

from degrading their ability to safely 

and competently perform their duties. 

In addition to the science of fatigue 

management, the NRC considered 

several factors in establishing the 

existing requirements. These additional 

factors were practical and safety 

considerations associated with the 

management of refueling outages for 

large LWRs, including the following: (1) 

the typical duration and frequency of 

outages; (2) the availability of contract 

personnel to perform the work; (3) the 

risk presented by the outage work while 

the reactor is shut down; and (4) the 

controls applied to the work that may 

limit the potential for latent errors to 

challenge reactor safety when the 

reactor is returned to power. The details 

of such considerations may differ for 

new reactor technologies or designs. 

Such considerations may not be relevant 

for some reactor designs (e.g., reactors 

capable of on-line refueling) and there 

may be additional, more pertinent 

factors to consider for other designs. 

The NRC is seeking stakeholder input 

on whether alternative fatigue 

management requirements applicable to 

outages should be adopted to support 

technology-inclusive approaches that 

would be appropriate to support the 

licensing and regulation of future 

commercial nuclear plants. Please 

provide your considerations and 

rationale for your recommendation. 

Part 26—Draft Regulatory Guidance 

Approach for Fatigue Management 

In support of this proposed rule, the 

NRC has issued DG–5078, ‘‘Fatigue 

Management for Nuclear Power Plant 

Personnel at Commercial Nuclear Plants 

Licensed Under 10 CFR part 53.’’ This 

DG describes methods the NRC staff 

considers acceptable for addressing 

certain aspects of FFD programs at 

commercial nuclear facilities licensed 

under part53. 

The NRC staff also intends to 

eventually transition this draft guide 

into an update to RG 5.73, ‘‘Fatigue 

Management for Nuclear Power Plant 

Personnel,’’ or the development of a 

new RG. At this point, NRC staff is 

considering four options for future RG 

development: 

• Option 1: Amend the existing RG. 

The NRC may develop an updated 

version of RG 5.73 that continues to 

endorse (with clarifications, additions, 

and exceptions) the guidance contained 

in NEI 06–11, ‘‘Managing Personnel 

Fatigue at Nuclear Power Reactor Sites,’’ 

Revision 1, and incorporates the topics 

discussed within DG–5078 as new NRC 

staff positions in section C of RG 5.73. 

• Option 2: Issue a new RG specific 

to part 53 licensees. The NRC may 

develop an entirely new RG applicable 

specifically to facilities licensed under 

part 53. This new RG would capture the 

guidance contained in DG–5078 and 

incorporate existing guidance (e.g., 

selected guidance in RG 5.73 and NEI 

06–11) that is considered to be 

technology inclusive in nature. The 

existing guidance (i.e., RG 5.73) would 

remain in place as the guidance for 

facilities licensed under parts 50 and 52. 

• Option 3: Review and potentially 

endorse new or revised industrydeveloped guidance. The NRC may 

engage with the industry regarding a 

potential update to industry guidance 

document NEI 06–11 or the 

development of new, separate industrydeveloped guidance specific to facilities 

licensed under part 53. The NRC would 

then review the new or revised 

industry-developed guidance within the 

NRC’s RG process, which includes 

opportunities for public participation. 

New or revised industry-developed 

guidance could incorporate DG–5078 or 

propose alternatives for the NRC to 

consider. 

• Option 4: Develop a comprehensive 

revision of the existing RG. The NRC 

may develop a more comprehensive 

revision of RG 5.73 that would 

explicitly detail all NRC positions 

reflected in the existing RG (including 

those endorsed positions currently 

contained in NEI 06–11, Revision 1), 

along with the guidance of DG–5078. 

Such a revision would thereby be a 

‘‘stand-alone’’ document, without 

reference to or explicit endorsement of 

separate, industry-developed guidance. 

The NRC is seeking stakeholder input 

regarding which of the four options 

listed above would be optimal (or 

whether there are other options that the 

NRC should consider). Please provide 

your considerations and rationale for 

your recommendation. 

Part 53—Overall Organization 

Part 53 is structured as one framework 

with subparts providing technical, 

licensing, and administrative 

requirements for the various stages of 

the life cycle of a commercial nuclear 

plant. The organization of part 53 in this 

manner puts a complete set of 

requirements for each stage of the life 

cycle in a separate subpart with 

additional subparts for licensing and 

administrative requirements. 

The NRC is seeking comment on the 

proposed organization of the 

requirements in part 53 and possible 

improvements to how specific 

requirements (e.g., examples of which 

specific sections) could be consolidated 

or otherwise reorganized to make the 

rule clearer or more concise. 

There are numerous references in 

proposed part 53 to other NRC 

regulations. Examples of such references 

include those in proposed § 53.610 to 

NRC regulations related to radiation 

protection (part 20), FFD (part 26), 

physical security (part 73), and MC&A 

(10 CFR part 74, ‘‘Material Control and 

Accounting of Special Nuclear 

Material’’) for facilities receiving 

byproduct or SNMs. 

The NRC is seeking comment on 

whether such references to other 

regulations in various sections in the 

proposed part 53 provide benefits to 

applicants and licensees, or to other 

stakeholders seeking to understand the 

regulatory framework under part 53, or 

whether such references could be 

removed to reduce the length of part 53. 

Part 53, Subpart B—Comprehensive 

Risk Metrics 

The NRC is proposing to require the 

use of comprehensive risk metrics and 

associated risk performance objectives 

as one of several performance standards 

in part 53. Comprehensive risk metrics 

could include a risk metric or set of risk 

metrics that approximate the total 

overall risk from the facility to the 


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extent practicable. Associated risk 

performance objectives are 

preestablished values indicative of the 

comprehensive risk metrics that are 

used during risk-informed decisionmaking to gauge plant safety. 

Specifically, comprehensive risk metrics 

and associated risk performance 

objectives would provide one element of 

the safety criteria for LBEs other than 

DBAs in the proposed § 53.220. 

Comprehensive risk metrics, in the form 

of the IEFR and the ILCFR, and 

associated risk performance objectives, 

in the form of the QHOs of 5×10¥7 per 

year and 2×10¥6 per year, respectively, 

were similarly used in the LMP 

methodology to ensure that other 

evaluation criteria were conservatively 

defined and as a tool for focusing 

attention on matters important to 

managing the risks posed by nuclear 

power plants. The use of such 

comprehensive risk metrics and 

associated risk performance objectives 

in an integrated risk-informed decisionmaking process is similar to that used in 

RG 1.174, ‘‘An Approach for Using 

Probabilistic Risk Assessment in RiskInformed Decisions on Plant-Specific 

Changes to the Licensing Basis,’’ 

Revision 3. 

The NRC is seeking comment on the 

use of comprehensive risk metrics and 

associated risk performance objectives 

in part 53 as one of several performance 

standards. The IEFR and ILCFR and the 

QHOs represent comprehensive risk 

metrics and associated risk performance 

objectives that the NRC has used for 

decades in a variety of capacities. What 

other performance standards could be 

used to address the comprehensive risks 

posed by proposed commercial nuclear 

plants? Please provide your 

considerations and rationale for your 

recommendation. 

If an applicant proposes a novel 

approach to comprehensive plant risk 

and the NRC approves the approach, 

should the resulting NRC-approved 

comprehensive plant risk metrics and 

associated risk performance objectives 

be codified or otherwise memorialized 

over time and, if so, how? 

Part 53, Subpart B—Defense in Depth 

Proposed § 53.250 would establish 

requirements based on the longstanding 

NRC philosophy of providing defense in 

depth to address uncertainties 

concerning the design, operation, and 

performance of commercial nuclear 

plants during LBEs. 

The NRC is seeking comment on the 

inclusion of the proposed requirements 

to assess and provide defense in depth. 

The NRC is also seeking comment on 

whether to include specific provisions 

in § 53.250 and subpart B to more 

explicitly address the possible role of 

inherent characteristics of some SSCs in 

preventing or mitigating unplanned 

events. The proposed § 53.250 is 

worded to preclude relying on a single 

engineered design feature to address the 

range of LBEs other than DBAs, which 

could possibly allow crediting inherent 

characteristics without further lines of 

defense. How could possible inherent 

characteristics of SSCs be considered in 

the proposed requirements in § 53.250 

or in any alternative requirements for 

defense in depth provided in response 

to this item? Please provide your 

considerations and rationale for your 

recommendation. 

Part 53, Subpart C—Probabilistic Risk 

Assessment 

Current consensus PRA standards 

provide processes for appropriately 

defining the scope of a PRA and 

determining applicability of supporting 

requirements to suit the specific needs 

of a given applicant under proposed 

part 53. In addition to assessing other 

aspects of PRA acceptability such as 

PRA peer reviews, NRC determinations 

of the acceptability of such PRAs would 

assess the appropriateness of the 

applicant-defined scope as part of 

determining the applicability of a 

consensus PRA standard supporting an 

application. This approach is consistent 

with the current state of practice and 

offers appropriate flexibility for PRAs to 

be developed and assessed based on the 

application they are used to support, 

which includes consideration of how 

PRA results and insights are relied 

upon, together with factors such as 

safety margin, simplicity of design, and 

treatment of uncertainty. 

The NRC is seeking comment on what 

additional guidance, if any, is needed 

regarding PRA acceptability for Part 53 

applicants and licensees. 

Part 53, Subparts C and D—Earthquake 

Engineering 

Proposed § 53.480 would establish 

requirements related to seismic design 

considerations. This proposed section is 

intended to provide a clear connection 

between siting activities and seismic 

design activities and to support various 

approaches to presenting seismic 

hazards and addressing those hazards in 

designs. The proposed requirements are 

intended to provide sufficient flexibility 

to allow approaches like those currently 

in parts 50 and 100 or approaches that 

might be endorsed by the NRC in the 

future that could incorporate more risk 

insights from PRAs. 

The NRC is seeking comment on 

whether the proposed requirements for 

earthquake engineering provide 

appropriate flexibility in addressing 

seismic risks while also ensuring that 

the regulations continue to adequately 

address seismic hazards. Please provide 

your considerations and rationale for 

your recommendation. 

Part 53, Subpart E—Construction and 

Manufacturing 

1. Proposed § 53.610(b)(1)(iii) would 

require procedures that describe how 

construction will be controlled so as not 

to impact other features important to the 

design (e.g., dewatering, slope stability, 

backfill, compaction, and seepage). 

The NRC is seeking comment on 

whether such specific requirements are 

useful or whether these requirements 

could be met through other 

requirements proposed in part 53 or 

already present in other relevant 

regulations (e.g., quality assurance 

requirements in appendix B to part 50). 

Part 53, Subparts E and H— 

Manufacturing Licenses 

1. The proposed requirements 

governing manufacturing are set forth in 

subpart E, and the proposed 

requirements governing the licensing 

processes are contained in subpart H. 

Some of the proposed requirements, 

including provisions related to the 

loading of unirradiated fuel into a 

manufactured reactor, are intended to 

cover a factory-fabrication model that 

has been suggested for some microreactor designs. However, as written, the 

proposed provisions are not limited to 

any size or type of reactor. 

The NRC is seeking comment on 

whether the proposed regulations are 

sufficient to govern various scenarios for 

the possible manufacturing and 

deployment of manufactured reactors. 

If a comment indicates that the 

proposed regulations are not sufficient, 

please describe the reasons why, 

including, if applicable, any plausible 

scenario for which the commenter 

believes the proposed regulations are 

not sufficient. 

2. The proposed regulations in 

subpart H allow holders of or applicants 

for a COL to reference an ML but do not 

include such a provision for the holder 

of or applicant for a CP or OL. This 

proposed change from the current 

relationship between subparts in part 52 

and the part 50 licensing process was 

made to simplify the provisions in the 

proposed part 53 for licensing and 

deploying manufactured reactors. 

The NRC seeks comment on whether 

part 53 should include provisions for an 

applicant for or a holder of a CP or an 

OL to reference an ML and, if so, how 

this should be done. 


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3. Proposed § 53.1295 states that the 

holder of an ML could not begin 

manufacture of a manufactured reactor 

less than 6 months before the expiration 

of the license. This limitation is similar 

to the current restriction in § 52.177, 

which states that the manufacture of a 

reactor cannot begin less than 3 years 

before the expiration of the license. The 

restriction was revised from 3 years in 

part 52 to 6 months in the proposed part 

53 in recognition of the likely use of 

MLs for a factory-fabrication model for 

micro-reactors. 

The NRC seeks comment on whether 

it is necessary or appropriate to revise 

the 3-year restriction in part 52 on when 

manufacturing activities could begin in 

relation to license expiration and, if so, 

what that restriction should be. 

4. Proposed § 53.1288 provides the 

finality provisions for MLs and 

includes, as does existing § 52.171, 

limitations on the NRC’s imposition of 

new requirements on either the design 

or the requirements for the manufacture 

of a manufactured reactor. No MLs have 

been issued under part 52 and there is 

no practical experience with the 

proposed finality sections. While the 

implications of the finality provisions 

related to the design of a manufactured 

reactor can reasonably be inferred from 

experience with DCs and COLs, there is 

no experience or available guidance 

regarding finality for ‘‘requirements for 

the manufacture of the manufactured 

reactor.’’ 

The NRC is seeking comment on the 

proposed finality provisions for MLs 

and specifically if and how finality for 

manufacturing processes might be 

requested and used. 

5. The NRC is seeking comment on 

the proposed regulations for the loading 

of fresh (unirradiated) fuel into a 

manufactured reactor for subsequent 

transport to a site for which the 

Commission has issued a COL that 

authorizes construction and operation of 

a commercial nuclear plant using the 

manufactured reactor. The proposed 

regulation includes provisions for 

loading of fuel into manufactured 

reactors at a manufacturing facility prior 

to transporting the fueled reactor to its 

deployment site, as suggested by some 

stakeholders. The NRC has historically 

viewed reactor operation as including 

fuel load, and existing NRC regulations 

reflect this view. While the Act 

authorizes the NRC to issue licenses to 

manufacture production or utilization 

facilities, it does not contain specific 

provisions on fueling or operating 

facilities licensed under an ML, and 

existing ML regulations under part 52 

do not include provisions for fuel load. 

The proposed rule addresses this 

matter by allowing an applicant to 

combine an ML with a part 70 license, 

which would authorize possession of a 

manufactured reactor in which the 

licensee has loaded unirradiated fuel 

provided at least two independent 

criticality prevention mechanisms are in 

place, each of which is sufficient to 

prevent criticality assuming optimum 

neutron moderation and neutron 

reflection conditions. This requirement 

would limit the possibility of creating 

fission products and allow the control of 

SNM, so that the loading of the fuel into 

a manufactured reactor could be 

governed primarily via a part 70 license 

and associated regulations (including 

those in subpart H of part 70). 

A specific topic on which the NRC is 

seeking comment is on the potential 

benefits of and issues with including the 

requirements of subpart H of part 70 

within the proposed regulations for 

loading fuel into manufactured reactors 

at the manufacturing facility. For 

example, should the NRC include a 

threshold for including the 

requirements of subpart H of part 70 

and, if so, what factors and decision 

criteria should be considered in such a 

threshold? If a comment indicates that 

the proposed regulations are not 

sufficient, please describe the reasons 

why, including the plausible scenarios 

for which the proposed regulations 

would not work or could be made to 

work better. 

6. Section 170, ‘‘Indemnification and 

Limitation of Liability,’’ of the Act states 

that each license under section 103 shall 

have as a condition of the license a 

requirement that the licensee have and 

maintain financial protection of such 

type and in such amounts as the NRC 

shall require. 

The NRC is seeking comment on 

whether the proposed regulations 

should include amounts of required 

financial protections for MLs for fueled 

manufactured reactors, and, if so, what 

would be appropriate amounts of 

required financial protection. 

7. Some stakeholders have suggested 

that a fueled manufactured reactor with 

appropriate protections against 

criticality should not be categorized as 

a utilization facility under NRC 

regulations or Section 11cc. of the Act. 

The NRC is seeking comment on 

possible approaches where the NRC 

could find that a fueled manufactured 

reactor would not be a utilization 

facility, the basis for such a finding, and 

the potential benefits of and potential 

issues with such a finding. 

8. Proposed requirement 

§ 53.620(d)(2)(i) would require a 

security program, including a physical 

security plan, for any ML authorizing 

possession of a manufactured reactor 

into which fuel has been loaded at the 

manufacturing facility. Currently, 

requirements in § 73.67(c)(1) only 

require that a physical security plan be 

submitted for those licensees who 

possess, use, transport, or deliver to a 

carrier for transport SNM of moderate 

strategic significance, or 10 kg or more 

of SNM of low strategic significance. 

The NRC is seeking comment on 

whether the proposed requirement: (1) 

should be specific to the facility type 

(i.e., manufacturing facility) or be 

specific to the category of material being 

used at the facility; (2) should apply to 

all manufacturing plants, including 

those at which licensees may only 

possess SNM of low strategic 

significance (i.e., category III), or only 

those facilities for which an applicant 

must submit a physical security plan 

per § 73.67(c)(1); or (3) should include 

more specific requirements on the 

supplemental security measures that 

may be needed for licensees possessing 

SNM of moderate strategic significance 

(i.e., category II)? 

9. Proposed requirement 

§ 53.620(d)(2)(i) would require a 

cybersecurity program. The proposed 

general cybersecurity performance 

requirements would be to provide 

reasonable assurance that a cyberattack 

could not adversely impact the 

functions performed by digital assets 

used by the licensee for implementing 

the physical security, radiation 

monitoring, and criticality 

requirements. 

The NRC is seeking comment on the 

following: (1) to what extent 

stakeholders envision physical security 

controls, radiation monitoring, and 

criticality controls at a manufacturing 

facility being digital; (2) to what extent 

should the ML holder be required to 

protect digital computer and 

communications systems that impact 

safety and security functions from a 

cyberattack at a manufacturing facility 

authorized to load fuel; and (3) whether 

the rule provides sufficient clarity on 

the cybersecurity measures needed for 

license issuance or if additional detail 

should be included either in the rule or 

in guidance? 

10. Proposed requirement 

§ 53.620(d)(2)(i)(B) would require that 

the physical security program be 

designed to prevent unintended and 

uncontrolled criticality events. This 

would include criticality events that are 

initiated maliciously. 

The NRC is seeking comment on 

whether the ML holder should be 

required to design its security program 

to protect against radiological sabotage 


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(i.e., an unintended criticality event 

leading to unacceptable radiological 

consequences), in addition to theft and 

diversion. For example, should the NRC 

establish security requirements to 

prevent an adversary, including an 

insider, from tampering with the reactor 

at a manufacturing facility or during 

transport in such a way as to cause an 

inadvertent criticality event? If so, 

should the NRC consider factors such as 

the category of fuel and the number of 

reactors at a factory that can 

simultaneously be loaded with fuel in 

establishing the security requirements? 

11. Proposed requirement 

§ 53.620(d)(2)(i) would require an ML 

holder to meet the performance 

objectives in § 73.67. Requirements 

§ 73.67(e) and § 73.67(g) include 

provisions for security of category II and 

category III quantities of SNM, 

respectively, during transportation. 

The NRC is seeking comment on the 

extent to which the ML should require 

ASMs (i.e., security measures above 

those required by § 73.67(e) and 

§ 73.67(g)) for transportation of a fueled 

reactor to its place of operation. What 

should those measures be? 

12. Proposed requirement 

§ 53.620(d)(2)(i) would require an ML 

holder to meet the performance 

objectives of § 73.67. For licensees 

utilizing a category II quantity of SNM, 

the requirement in § 73.67(d)(4) would 

have the ML holder conduct a screening 

to confirm the identity of an individual 

prior to granting unescorted access to 

the controlled access area where the 

material is used or stored. The purpose 

of this requirement is to both confirm 

the identity of the individual and 

support a determination that the 

individual is trustworthy and reliable. 

The NRC is seeking comment on 

whether the ML requirements should 

include ASMs (i.e., measures beyond 

those required by § 73.67(d)(4)) in order 

to provide reasonable assurance of 

identity confirmation and 

trustworthiness and reliability. 

13. The NRC is seeking comment on 

whether provisions regulating the 

testing of fueled manufactured reactors 

in the manufacturing facility should be 

included in part 53 and, if so, what 

would be practical for the holder of an 

ML while also providing adequate 

protection of public health and safety. 

One possibility could be COLs that 

would be issued to the holders of an ML 

to cover low power (e.g., <5% rated 

thermal power) nuclear physics testing 

of fueled manufactured reactors within 

the manufacturing facility prior to the 

manufactured reactors being transported 

to and incorporated into a commercial 

nuclear plant for the purpose of energy 

production. The NRC recognizes 

configuration changes are needed to 

perform nuclear physics testing and is 

seeking comment on what requirements 

should apply to the manufactured 

reactors and the manufacturing facility 

during such testing (e.g., limiting power 

levels). If a comment indicates that the 

regulations should address limited 

operations at manufacturing facilities, 

please describe the likely scenarios that 

would need to be addressed and suggest 

what would be appropriate 

requirements for such scenarios. 

While an ML holder could 

accomplish nuclear physics testing by 

applying for a COL under the proposed 

subpart H of part 53, stakeholders have 

indicated that many of the requirements 

would likely be unnecessary, given the 

reduced risk profile posed by such 

activities. Therefore, the NRC is seeking 

comment on what requirements in 

subpart H of part 53 should apply to 

applicants for a COL who would 

perform testing of fueled manufactured 

reactors at the manufacturing plant. 

Examples of proposed requirements that 

might be relaxed or modified for 

applications for low power testing at 

manufacturing plants include those 

related to selection of LBEs to reflect 

limited inventory of radionuclides and 

decay heat, aircraft impact assessments, 

and earthquake engineering. 

Additionally, the NRC is seeking 

comment on whether several other 

requirements in part 53 could be 

modified for applications for a low 

power testing COL at a manufacturing 

facility. For example, the NRC is seeking 

comment on how portions of the ML 

facility used to support testing should 

fall within the requirements for 

construction activities under § 53.610; 

whether §§ 53.710 and 53.715 (SSC 

configuration control) must be 

implemented to ensure portions of the 

ML facility relied on to limit potential 

radiological consequences from LBEs 

are available to perform their safety 

functions; and whether the 

requirements of § 53.730 could be 

modified to reflect the conditions of low 

power physics testing. If a comment 

indicates that some design and analysis 

requirements and related application 

requirements in subpart H of the 

proposed part 53 are not needed for the 

testing of fueled manufactured reactors, 

please provide a rationale supporting 

your comment and, if applicable, what 

alternate requirements would be 

appropriate. 

Moreover, the licensing mechanism 

for the facility could present unique 

challenges. One option could be to issue 

a low power testing COL for each fueled 

manufactured reactor to be tested. This 

would comport with the agency’s 

practice of issuing one license per 

reactor but could prove prohibitive from 

a cost standpoint and may provide very 

little safety benefit if all manufactured 

reactors are the same. Alternatively, one 

low power testing COL could be issued 

for the portions of the ML facility used 

to test the fueled manufactured reactors 

and allow multiple fueled manufactured 

reactors to be completed and tested over 

the course of the ML. Under this 

approach, any ITAAC related to testing 

of the fueled manufactured reactors 

would need to be closed after they were 

manufactured but prior to testing, and 

the NRC would issue a notice of 

intended operation and provide the 

public an opportunity to request a 

hearing on whether each fueled 

manufactured reactor as constructed 

complies, or on completion will 

comply, with the acceptance criteria of 

the license. The NRC is seeking 

comment on the potential benefits and 

issues with having a COL for each 

fueled manufactured reactor to be tested 

versus having a COL cover the testing of 

multiple fueled manufactured reactors. 

If a comment indicates a preference for 

a particular approach, please provide a 

rationale supporting the comment and 

describe the specific scenarios that the 

regulations need to address. 

Part 53, Subpart F—Staffing and 

Generally Licensed Reactor Operators 

Under the Act Sections 106 and 107, 

the NRC is proposing to group 

commercial reactors into classes upon 

the basis of the similarity of operating 

and technical characteristics of the 

facilities, and then to prescribe uniform 

conditions for licensing individuals as 

operators of any of the various classes; 

determine the qualifications of such 

individuals; and, for certain classes of 

commercial reactors, issue general 

licenses (i.e., licenses for which no 

application is needed) to such 

individuals allowing the individuals to 

operate the commercial reactor. 

1. Categories of Individuals Who May 

Manipulate Facility Controls: The NRC 

is proposing requirements that would 

allow the manipulation of the controls 

of certain facilities by GLROs in lieu of 

specifically licensed reactor operators 

and senior reactor operators. Reactor 

operators and senior reactor operators 

are the only categories of individuals 

currently allowed to be licensed to 

manipulate the controls of utilization 

facilities under part 55. 

The NRC is interested in public 

perspectives on this proposed addition 

of the GLRO category, particularly in 

light of new reactor technologies and 

concepts of operations. 


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2. Criteria for GLRO Staffing: The 

NRC is proposing criteria under which 

facilities would be staffed by GLROs in 

lieu of specifically licensed reactor 

operators and senior reactor operators. 

These criteria establish a new class of 

self-reliant-mitigation facilities, as 

defined in part 53, for which distinct 

GLRO licensing and staffing 

requirements would apply. 

The NRC is soliciting public feedback 

regarding whether these proposed 

criteria are appropriate and what, if any, 

alternative criteria should be 

considered. Please provide your 

considerations and rationale for your 

answer. 

3. Medical Requirements for GLROs: 

Based on the proposed criteria that a 

self-reliant-mitigation facility, as 

defined in part 53, must meet, the NRC 

is proposing not to subject GLROs to 

requirements for medical fitness and 

medical examination. This is in contrast 

with the proposed requirements 

associated with specifically licensed 

reactor operators and senior reactor 

operators, as well as the existing 

requirements for reactor operators and 

senior reactor operators under part 55. 

The NRC is soliciting public feedback 

regarding whether GLROs should be 

subject to medical fitness and/or 

medical examination requirements like 

reactor operators and senior reactor 

operators. Please provide your 

considerations and rationale for your 

answer. 

4. Onshift Engineering Expertise: The 

NRC is proposing to require that 

engineering expertise be accounted for 

within facility staffing plans. This 

proposed requirement would be in lieu 

of the traditional position of the Shift 

Technical Advisor. The NRC is further 

proposing that individuals providing 

such engineering expertise would need, 

among other things, to possess either a 

qualifying 4-year degree or licensure as 

a Professional Engineer. 

The NRC is interested in feedback 

from the public regarding the 

appropriateness of this requirement, 

including any alternatives that should 

be considered. Please provide your 

considerations and rationale for your 

answer. 

5. Use of Simulation Facilities as HFE 

Testbeds: The NRC is proposing to 

establish regulations pertaining to the 

use of simulation facilities within the 

context of the licensing programs both 

for specifically licensed reactor 

operators and senior reactor operators as 

well as for GLROs. However, these 

regulations, as currently proposed, do 

not address the use of simulation 

facilities within the context of serving as 

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 § 50.160, ‘‘Emergency 

preparedness for small modular 

reactors, non-light-water reactors, and 

non-power production or utilization 

facilities,’’ and by making conforming 

changes within § 50.160. The proposed 

framework for part 53 would also 

introduce a graded approach to physical 

protection requirements that includes 

the criterion in § 53.860(a)(2)(i) to 

establish a class of licensees that would 

not be required to protect against the 

design-basis threat (DBT) of radiological 

sabotage. The NRC is soliciting public 

comment relating to these topics, which 

could include ways that graded 

approaches for both emergency 

preparedness and security programs 

might be assessed and considered 

during the licensing process. 

The NRC is seeking comment on the 

sufficiency and clarity of requirements 

in proposed part 53 related to the 

assessments needed to support graded 

emergency planning and security. If a 

comment indicates that there is an issue 

with the sufficiency or clarity of the 

proposed regulations, please describe 

the reasons why, including, if 

applicable, any scenario for which the 

proposed regulations are not sufficient 

and possible ways to clarify the 

requirements. The NRC is specifically 

seeking comment on possible challenges 

arising from the interactions between 

the proposed regulations and related 

assessments for grading the 

requirements for emergency planning 

and security. 

2. The NRC is preparing various 

guidance documents to support this 

rulemaking and other ongoing or 

recently completed rulemakings related 

to emergency preparedness and 

security. DG–5076, ‘‘Guidance for 

Technology-Inclusive Requirements for 

Physical Protection of Licensed 

Activities at Commercial Nuclear 

Plants,’’ has been issued along with this 

proposed rulemaking and public 

comments are requested via this notice 

on that draft guidance. The NRC is also 

planning to issue a draft revision of RG 

1.242, ‘‘Performance-Based Emergency 

Preparedness for Small Modular 

Reactors, Non-Light-Water Reactors, and 

Non-Power Production or Utilization 

Facilities,’’ for public comment. The 

planned revision to RG 1.242 would add 

guidance for part 53 applicants and 

licensees. 

In the staff requirements 

memorandum to SECY–23–0021, the 

Commission directed the NRC staff to 

address the consideration of securityrelated events for an advanced reactor 

that addresses security through design 

and engineered safety features when it 

harmonizes this rulemaking with the EP 

for SMR/ONT rule. In the EP for SMR/ 

ONT rule, the NRC established an 

alternative performance-based and riskinformed approach for emergency 

planning, including determining the 

need for and size of an emergency 

planning zone (EPZ) to support 

predetermined, prompt protective 

actions. The NRC has incorporated the 

relevant rule language from the EP for 

SMR/ONT rule into this proposed rule 

and is seeking stakeholder feedback as 

to whether additional rule language 

changes or additional guidance would 

be beneficial. 

In light of the Commission direction 

and the above considerations, the NRC 

is assessing how best to address the 

treatment of security-related events in 

emergency planning, including in the 

determination of EPZ size, for reactors 

licensed under part 53. Part 53 is 

introducing an alternative approach to 

meeting security regulations that should 

be taken into consideration under 

§ 50.160. Stakeholders are encouraged to 

take a holistic view of the various 

activities and opportunities to provide 

comments on this rulemaking and 

related guidance supporting this 

rulemaking (e.g., DG–5076 on physical 

protection requirements, future 

revisions to RG 1.242). In developing 

comments, the NRC urges stakeholders 

to consider various scenarios that might 

arise when implementing graded 

approaches for security and emergency 

planning for various reactor designs. 

Scenarios could include the following: 

• the potential consequences from 

security events up to and including the 

DBT of radiological sabotage are 

bounded by unlikely and very unlikely 

event sequences such that security 

events do not need separate analyses in 

the EPZ size determination; 


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• the potential consequences from 

security events up to and including the 

DBT are not bounded by unlikely and 

very unlikely event sequences but could 

otherwise support a reduced EPZ size 

consistent with considerations 

discussed in RG 1.242 and NUREG– 

0396, ‘‘Planning Basis for the 

Development of State and Local 

Government Radiological Emergency 

Response Plans in Support of Light 

Water Nuclear Power Plants’’; or 

• the potential consequences from 

security events up to and including the 

DBT are not bounded by unlikely and 

very unlikely event sequences and 

warrant consideration of increasing the 

size of the EPZ. 

The NRC is interested in comments 

on the need for additional rule language 

or guidance to address graded 

approaches for emergency planning and 

security programs under the scenarios 

described above for part 53 applicants 

and licensees. Please address within the 

comments any technical, policy, or legal 

issues that are associated with your 

suggestions. 

Part 53, Subpart F—Integrity 

Assessment Program Requirements 

Decades of operating experience with 

LWRs suggests that phenomena such as 

environmentally assisted fatigue and 

chemical interactions could impact 

certain SSCs during the life of a 

commercial nuclear plant. Under the 

existing regulatory framework, 

historically, some of these phenomena 

were not addressed during early 

licensing reviews but were identified 

and addressed later when significant 

safety issues arose (e.g., see numerous 

generic letters, bulletins, orders, and 

development and implementation of 

vessel integrity and materials reliability 

programs) or a licensee voluntarily 

pursued renewal of an OL under part 

54. The NRC is proposing to include a 

new set of programmatic requirements 

for an Integrity Assessment Program that 

would ensure these phenomena are 

addressed early in the life of a 

commercial nuclear plant licensed 

under part 53. The requirements would 

be provided in § 53.870. 

The NRC is seeking comment on 

whether the proposed requirements 

under the Integrity Assessment Program 

appropriately complement design 

requirements to address concerns 

regarding aging, cyclic or transient load 

limits, and degradation mechanisms 

related to chemical interactions, 

operating temperatures, effects of 

irradiation, and other environmental 

factors. In addition, the NRC is 

interested in views on whether, and if 

so how, degradation mechanisms are or 

could be addressed in other programs. 

Part 53, Subpart G—Decommissioning 

1. On March 3, 2022, the NRC 

published the proposed rule entitled 

‘‘Regulatory Improvements for 

Production and Utilization Facilities 

Transitioning to Decommissioning’’ (87 

FR 12254). This rulemaking would 

amend the NRC’s current regulations to 

provide an appropriate regulatory 

framework for nuclear power reactors 

transitioning from operations to 

decommissioning. The rulemaking 

would address lessons learned from 

licensees that have completed or are 

currently in the decommissioning 

process. The NRC staff sent a draft final 

rule to the Commission for its 

consideration on January 31, 2024, in 

SECY–24–0011, ‘‘Final Rule: Regulatory 

Improvements for Production and 

Utilization Facilities Transitioning to 

Decommissioning (3150–AJ59; NRC– 

2015–0070).’’ 

What aspects of this draft final rule, 

if any, should be incorporated in a part 

53 final rule and why? 

2. Proposed § 53.1060(b) in subpart G 

would require that, ‘‘No later than 30 

days after the Commission publishes 

notice in the Federal Register under 

§ 53.1452(a), the licensee must submit a 

report containing a certification that 

financial assurance for 

decommissioning is being provided in 

an amount specified in the licensee’s 

most recent updated certification, 

including a copy of the financial 

instrument obtained to satisfy 

§ 53.1040.’’ This is similar to the current 

requirement in § 50.75(e)(3) for part 52 

COL holders. The NRC is seeking 

comment on whether commercial 

nuclear plant COL holders under part 53 

should have the same requirement as 

COL holders under part 52 to 

demonstrate that they have financial 

assurance in place no later than 30 days 

after the Commission issues the notice 

of intended operation under § 53.1452. 

Please provide your considerations and 

rationale for your answer. 

Part 53, Subpart H—Licenses To 

Construct and Operate Commercial 

Nuclear Plants of Identical Design at 

Multiple Sites 

In addition to including provisions in 

part 53, subpart H, for referencing ESPs, 

standard design approvals, and design 

certifications in applications for 

commercial nuclear plants, the 

proposed § 53.1470 provides optional 

requirements related to the submittal 

and NRC review of CP, OL, and COL 

applications to construct and operate 

commercial nuclear plants of identical 

design at multiple sites, similar to 

requirements found in appendix N in 

both 10 CFR parts 50 and 52. This 

section would set out the particular 

requirements and provisions applicable 

to situations in which applications for 

CPs and subsequent OLs, or COLs, 

under this part, are filed by one or more 

applicants for licenses to construct and 

operate nuclear power reactors of 

identical design (‘‘common design’’) to 

be located at multiple sites. Hearings for 

applications filed under appendix N in 

both parts 50 and 52 are governed by 

subpart D of part 2, as would be the case 

for future part 53 applications under 

proposed § 53.1470. 

Under the proposed requirements in 

this section, each application is to be 

treated as a separate application, with 

the exception of the common design, 

and so would require separate 

applications, separate determinations of 

sufficiency for docketing, separate 

notices of docketing, and so forth. 

Proposed § 53.1470 would also require 

that each application list all the 

applications that are to be treated 

together to ensure that the NRC is 

clearly informed of the intentions of all 

applicants. Ordinarily, the NRC would 

publish in the Federal Register a 

separate notification of docketing for 

each application, so that delays in the 

docketing of one application would not 

delay the docketing and subsequent 

technical review of other applications. 

However, if circumstances allow (e.g., 

sufficiency review for multiple 

applications are completed 

simultaneously), the NRC could publish 

a single notice of docketing for multiple 

applications. 

With regard to how the NRC would 

fulfill its obligations under the National 

Environmental Policy Act of 1969, as 

amended, the NRC staff would prepare 

a separate environmental document for 

each application, but the NRC could 

conduct joint scoping on environmental 

issues related to the common design. If 

the applications reference a standard 

design certification or the use of a 

manufactured reactor, then the 

environmental document would need to 

incorporate by reference the 

environmental assessment (EA) 

prepared for either the design 

certification or the ML, as applicable. In 

addition, § 53.1470 would require the 

ACRS to report on each of the 

applications, as would be required by 

provisions in subpart H of part 53. Each 

ACRS report would be limited to the 

safety matters which are not relevant to 

the common design. In addition, the 

ACRS would need to issue a report on 

the safety of the common design— 

except for those matters relevant to the 


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15Section 70.23(a)(5). 

safety of a referenced design 

certification or manufactured reactor. 

Given this synopsis of how the 

requirements in proposed § 53.1470 

would be implemented as currently 

written, the NRC is seeking comment on 

whether there are opportunities to allow 

added flexibility for applicants under 

these provisions. This could include 

consideration of whether applications 

for which the ‘‘common design’’ is not 

completely identical could be evaluated 

under this provision and, if so, what the 

process would be for determining the 

appropriateness of a common review. In 

addition, the NRC is interested in 

feedback about the pros and cons of 

requiring that applications under these 

proposed provisions be submitted at the 

same time versus allowing them to be 

submitted on a staggered basis. 

Part 53, Subparts H and I—Probabilistic 

Risk Assessment Information 

Proposed § 53.1239(a)(18) in subpart 

H and the related references to this 

proposed requirement for the holders of 

OLs and COLs would require a 

description of the PRA required by 

§ 53.450(a), and its results to be 

included in FSARs. However, guidance 

documents may further clarify the 

division of PRA-related information 

needed to be in the FSAR, in other 

possible licensing basis documents, and 

controlled as plant records subject to 

inspections and audits. For example, a 

possible approach could be to include a 

summary of the PRA results in the 

FSAR and control that information 

under § 53.1545 and create a separate 

document related to the broader PRA 

analyses and related processes as a 

program document under § 53.1560. The 

program document would provide more 

detail than the summaries in the FSAR 

but still be a much-condensed source of 

information in comparison to the 

documentation of the PRA. This 

possible approach would reflect the role 

of the PRA in the licensing process 

under part 53 and in maintaining 

margins to the safety and evaluation 

criteria in subparts B and C but may 

allow a more appropriate evaluation 

process to address the particulars and 

complexities of the PRA-related 

documents. 

The NRC is seeking comment on the 

appropriate placement of PRA-related 

information among various licensing 

basis documents and plant records. In 

addition to the placement of PRArelated information, the NRC is seeking 

comment on the appropriate control of 

that information and on the routine 

submittal of updates to the NRC. Please 

provide your considerations and 

rationale for your answer. 

Part 53, Subparts H and I—Changes to 

Manufacturing Licenses 

Proposed § 53.1530 would not allow 

the holder of an ML or the holder of a 

COL using a manufactured reactor to 

make changes to the design of the 

manufactured reactor without 

requesting a license amendment from 

the NRC. The proposed requirements do 

not include a specific mention of the 

manufacturing processes for which the 

NRC could possibly provide finality 

under proposed § 53.1288. 

The NRC is seeking comment on the 

appropriate change control provisions 

for MLs, including whether criteria 

could be developed to determine when 

a license amendment request would not 

be required and whether those criteria 

should address changes in 

manufacturing processes as well as 

changes in the design. Please provide 

your considerations and rationale for 

your recommendation. 

Financial Qualifications 

Utility new reactor applicants are 

exempt under § 50.33(f) from financial 

qualification reviews because they are 

generically presumed to be financially 

qualified for operations. In contrast, 

merchant power plant new reactor 

applicants are required under 

§ 50.33(f)(2) to submit information that 

demonstrates they possess or have 

reasonable assurance of obtaining the 

funds necessary to cover estimated 

construction and operating costs for the 

period of the license. A ‘‘merchant 

power plant new reactor applicant’’ is a 

non-rate-regulated entity (e.g., a 

nonutility) that engages in the business 

of production, manufacturing, 

generating, buying, aggregating, 

marketing, or brokering electricity for 

sale at wholesale or for retail sale to the 

public. Over the past decade, the agency 

has heard some concerns about the 

challenges that merchant power plant 

applicants face in demonstrating 

compliance with the current financial 

qualification requirements. 

Does this standard continue to pose 

challenges for merchant power plant 

applicants? If so, please provide a 

detailed explanation of these challenges. 

Should part 53 have the same 

financial qualification requirements as 

parts 50 and 52? Why or why not? 

Are there categories of merchant new 

reactor applicants for which a part 70 

‘‘appears to be financially qualified’’ 

standard would be more appropriate? 15

If so, please explain what types of 

applicants should be able to use the part 

70 financial qualification standard and 

what distinguishes these applicants 

from ones that should not be able to use 

this standard. 

If a part 70 financial qualification 

standard were to apply to a category of 

merchant new reactor applicants, 

should it also apply to pre-construction 

license transfer applications for these 

reactors? Why or why not? 

Is there another standard the agency 

should consider for financial 

qualification of merchant new reactor 

applicants? Commenters are encouraged 

to provide specific suggestions and the 

basis for those suggestions. 

Part 73, Section 73.100—Physical 

Security 

The proposed § 73.100 would identify 

the proposed performance-based 

physical security requirements with 

which future commercial power reactor 

applicants or licensees’ physical 

protection programs would need to 

demonstrate compliance, without 

prescribing the specific methods that 

must be used to satisfy them. Applicants 

and licensees would have increased 

flexibility regarding the modern 

technologies and methods that they 

could use. Implementing guidance in 

DG–5076 (proposed RG 5.97), 

‘‘Guidance for Technology Inclusive 

Requirements for Physical Protection of 

Licensed Activities at Commercial 

Nuclear Plants,’’ would be available to 

assist applicants and licensees. For 

example, DG–5076 provides detailed 

guidance, including performance 

standard recommendations, on the 

probability of detection and alternative 

sources of power for exterior intrusion 

detection systems (subsection 4.1.1.1.A), 

interior intrusion detection (subsection 

4.1.1.1.B), intrusion assessment 

(subsection 4.1.1.2.A), security 

response/neutralization subsection 

(4.1.1.4.A), security communication 

(subsection 4.1.1.3.A), and security 

delay (subsection 4.1.1.4.C). 

Does the NRC’s proposed approach in 

§ 73.100 provide a sufficient level of 

detail to be readily understood and 

easily applied to the licensing and 

oversight of new and advanced power 

reactors, or should the NRC consider 

moving some objective and measurable 

security performance standard 

recommendations from the draft 

implementing guidance in DG–5076 

into proposed § 73.100? If so, which 

objective and measurable security 

performance standard recommendations 

should be moved from DG–5076 to 

§ 73.100? Please provide the basis for 

your response. 


lotter on DSK11XQN23PROD with PROPOSALS2
86988 Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules 

Part 73, Section 73.110—Cybersecurity 

The proposed § 73.110 would require 

licensees to demonstrate protection 

against cyberattacks in a manner that is 

commensurate with the potential 

consequences from those attacks, 

without prescribing the specific 

methods that must be used to 

demonstrate protection. Under proposed 

§ 73.110(a), licensees would need to 

ensure that digital computer and 

communications systems are adequately 

protected against a potential cyberattack 

that would, for example, result in 

adverse impacts to the physical security 

digital assets used by the licensee to 

prevent unauthorized removal of 

material per § 53.860(a). Protecting 

against such a potential cyberattack 

would involve requiring cybersecurity 

for SNM at a commercial nuclear reactor 

licensed under part 53. Applicants and 

licensees would have increased 

flexibility regarding the modern 

technologies and methods that they 

could use for protecting against such a 

potential cyberattack. Detailed 

implementing guidance in DG–5075 

(proposed RG 5.96), ‘‘Establishing 

Cybersecurity Programs for Commercial 

Nuclear Plants licensed under 10 CFR 

part 53,’’ would be available to assist 

applicants and licensees. For example, 

DG–5075 provides guidance on the 

implementation of security by design 

features (e.g., facility design) for 

negating the potential consequences 

from such a potential cyberattack. 

If a cyberattack were to compromise 

the availability, integrity, or 

confidentiality of data or systems 

associated with security systems/ 

measures for the protection of SNM at 

a commercial nuclear reactor licensed 

under part 53, do the potential 

consequences warrant requiring 

cybersecurity for such material? Please 

provide the basis for your response 

including a detailed explanation of 

challenges, if any, posed by requiring 

cybersecurity for SNM at a commercial 

nuclear reactor licensed under part 53. 

Recent Legislation 

On July 9, 2024, the President signed 

into law the Accelerating Deployment of 

Versatile, Advanced Nuclear for Clean 

Energy Act of 2024, also referred to as 

the ADVANCE Act. Section 203, 

‘‘Licensing Considerations Relating to 

Use of Nuclear Energy for Nonelectric 

Applications,’’ and Section 208, 

‘‘Regulatory Requirements for MicroReactors,’’ of the ADVANCE Act 

specifically mention the technologyinclusive regulatory framework to be 

established under section 103(a)(4) of 

NEIMA as a potential vehicle to be 

considered for the report to Congress 

required under section 203 and a 

potential vehicle to implement 

strategies and guidance for the licensing 

and regulation of micro-reactors 

required under section 208. This 

proposed rulemaking is, in part, how 

the NRC is implementing section 

103(a)(4) of NEIMA. 

The NRC is seeking comment on how 

part 53 could be revised to better enable 

its potential use to implement the 

ADVANCE Act. Specifically, Section 

208 of the ADVANCE Act requires the 

NRC to develop and implement ‘‘riskinformed and performance-based 

strategies and guidance’’ in several areas 

for the licensing and regulation of 

micro-reactors, including with respect 

to ‘‘licensing mobile deployment.’’ The 

ADVANCE Act requires the NRC to 

consider ‘‘the unique characteristics of 

micro-reactors,’’ including physical size, 

design simplicity, and source term; 

opportunities to incorporate specific 

improvements related to streamlining 

the review process; and other policy and 

licensing issues. With regard to 

implementation, the ADVANCE Act 

provides the NRC with three options. 

The NRC may implement the developed 

strategies and guidance, as appropriate, 

via (1) the existing regulatory 

framework, (2) the Part 53 rulemaking, 

or (3) a pending or new rulemaking. 

Given the language included in Section 

208, the NRC is seeking comment on 

how part 53 could be revised to better 

address the ADVANCE Act’s 

requirements related to strategies and 

guidance for micro-reactors. 

VII. Section-by-Section Analysis 

The following paragraphs describe the 

specific changes proposed by this 

rulemaking. 

§ 1.43 Office of Nuclear Reactor 

Regulation 

This proposed rule would revise 

§ 1.43(a)(2) to extend the authority of 

the Office of Nuclear Reactor Regulation 

to regulate source, byproduct, and SNM 

at facilities licensed under part 53. 

§ 2.1 Scope 

This proposed rule would revise 

§ 2.1(e) to apply to standard design 

approvals under part 53. 

§ 2.4 Definitions 

This proposed rule would revise § 2.4 

to update the definition of ‘‘Contested 

proceeding’’ to include NRC 

enforcement actions against applicants 

for a standard DC under part 53. It 

would also update the definition of 

‘‘Facility’’ to encompass utilization 

facilities as defined in § 53.020 (there 

are no production facilities under part 

53). 

§ 2.100 Scope of Subpart 

This proposed rule would revise 

§ 2.100 to extend the scope of subpart A 

to licenses and standard design 

approvals issued under §§ 53.1200 

through 53.1221. 

§ 2.101 Filing of Application 

This proposed rule would revise 

§ 2.101 to be applicable to part 53 

applicants in addition to part 50 and 52 

applicants by adding references to part 

53 in paragraphs (a)(3)(i), (a)(5), and 

(a)(9). 

§ 2.104 Notice of Hearing 

This proposed rule would extend the 

hearing notice requirement in § 2.104(a) 

to applications concerning facilities 

covered under part 53. Footnote 1 to 

§ 2.104 would be revised in a 

corresponding manner. 

§ 2.105 Notice of Proposed Action 

This proposed rule would revise 

§ 2.105 to extend the requirement in 

§ 2.104 to publish a notice of intended 

operation or a notice of proposed action, 

as applicable, to part 53 applicants in 

addition to part 50 and 52 applicants by 

adding corresponding references to part 

53 in paragraphs (a), (a)(4), (a)(10), 

(a)(12), (a)(13), and (b)(3). 

§ 2.106 Notice of Issuance 

This proposed rule would revise 

§ 2.106 to extend the issuance notice 

requirement to applications concerning 

facilities covered under part 53 through 

updated references in paragraphs (a)(2) 

and (3), and (b)(2). 

§ 2.109 Effect of Timely Renewal 

Application 

This proposed rule would revise 

§ 2.109 to add references to part 53 in 

paragraphs (b), (c), and (d) regarding the 

timing of license renewal applications. 

§ 2.110 Filing and Administrative 

Action on Submittals for Standard 

Design Approval or Early Review of Site 

Suitability Issues 

This proposed rule would revise 

§ 2.110 to include references to part 53 

in paragraphs (a)(1) and (b). 

§ 2.202 Orders 

This proposed rule would revise 

§ 2.202(e) to add references to part 53 

regarding the requirements to be 

followed for orders involving the 

modification of a license, COL, ESP, 

standard DC rule, standard design 

approval, or ML. 


lotter on DSK11XQN23PROD with PROPOSALS2
Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules 86989 

§ 2.309 Hearing Requests, Petitions To 

Intervene, Requirements for Standing, 

and Contentions 

This proposed rule would revise 

§ 2.309 to include references to part 53 

in paragraphs (a), (f)(1)(i), (f)(1)(vi) and 

(vii), (g), (h)(2), (i)(2), and (j) regarding 

a request for hearing under § 53.1452. 

§ 2.310 Selection of Hearing 

Procedures 

This proposed rule would revise 

§ 2.310 by revising paragraph (a), the 

introductory text for paragraph (h), and 

paragraphs (i) and (j) to incorporate 

references to part 53 regarding hearing 

procedures. 

§ 2.329 Prehearing Conference 

This proposed rule would revise 

§ 2.329(a) to extend the timing 

requirements for prehearing conferences 

involving CPs and licenses under part 

53. 

§ 2.339 Expedited Decision-Making 

Procedure 

This proposed rule would revise 

§ 2.339(d) to include references to part 

53 regarding expedited decision-making 

procedures. 

§ 2.340 Initial Decision in Certain 

Contested Proceedings; Immediate 

Effectiveness of Initial Decisions; 

Issuance of Authorizations, Permits and 

Licenses 

This proposed rule would revise 

§ 2.340 regarding initial decisions of a 

presiding officer in certain contested 

proceedings, the effective date of those 

decisions, and the issuance of 

authorizations, permits, and licenses, by 

incorporating references to part 53 in 

paragraphs (b), (c), (d), (f), (i), and (j). 

§ 2.341 Review of Decisions and 

Actions of a Presiding Officer 

This proposed rule would revise 

§ 2.341(a)(1) to include an updated 

reference to part 53 regarding the 

allowance of a period of interim 

operation. 

§ 2.400 Scope of Subpart 

This proposed rule would revise 

§ 2.400 to extend the scope of subpart D 

of part 2 to include part 53 applicants 

for licenses to construct or operate 

nuclear power reactors of identical 

design at multiple sites. 

§ 2.401 Notice of Hearing on 

Construction Permit or Combined 

License Applications Pursuant to 

Appendix N of 10 CFR Parts 50, 52, or 

53 

This proposed rule would revise the 

section heading and § 2.401 to extend 

the hearing notice requirement to 

applications concerning facilities 

covered under part 53. 

§ 2.402 Separate Hearings on Separate 

Issues; Consolidation of Proceedings 

This proposed rule would revise 

§ 2.402(a) to apply provisions regarding 

separate hearings and the consolidation 

of proceedings to part 53 applicants. 

§ 2.403 Notice of Proposed Action on 

Applications for Operating Licenses 

Pursuant To Appendix N of 10 CFR Part 

50 

This proposed rule would revise 

§ 2.403 to require the Commission to 

publish a notification of proposed 

action in the Federal Register after 

applications under part 53 are docketed. 

§ 2.404 Hearings on Applications for 

Operating Licenses Pursuant to 

Appendix N of 10 CFR Part 50 

This proposed rule would revise 

§ 2.404 to apply to applications for an 

OL under part 53. 

§ 2.405 Initial Decisions in 

Consolidated Hearings 

This proposed rule would revise 

§ 2.405 to be applicable to CPs, fullpower OLs, and COLs under part 53. 

§ 2.406 Finality of Decisions on 

Separate Issues 

This proposed rule would revise 

§ 2.406 to be applicable to proceedings 

conducted pursuant to part 53. 

§ 2.500 Scope of Subpart 

This proposed rule would revise 

§ 2.500 to extend the provisions of 

subpart E of part 2 to include 

applications for a license to 

manufacture nuclear power reactors 

under part 53. 

§ 2.501 Notice of Hearing on 

Application Under Subpart F of 10 CFR 

Part 52 or 53 for a License To 

Manufacture Nuclear Power Reactors 

This proposed rule would revise the 

section heading and § 2.501(a) by 

extending its provisions to applications 

for a license to manufacture nuclear 

power reactors under part 53. 

§ 2.643 Acceptance and Docketing of 

Application for Limited Work 

Authorization 

This proposed rule would revise 

§ 2.643(b) regarding the acceptance and 

docketing of an application for a CP for 

a utilization facility of the type specified 

in part 53. 

§ 2.645 Notice of Hearing 

This proposed rule would revise 

§ 2.645(a) to incorporate a reference to 

part 53. 

§ 2.649 Partial Decisions on Limited 

Work Authorization 

This proposed rule would revise 

§ 2.649 to extend its provisions to LWAs 

issued under part 53. 

§ 2.800 Scope and Applicability 

This proposed rule would revise 

§ 2.800 by revising paragraphs (c) and 

(d) to incorporate references to part 53 

regarding the scope and applicability of 

the rulemaking procedures contained in 

this subpart. 

§ 2.801 Initiation of Rulemaking 

This proposed rule would revise 

§ 2.801 to include a reference to part 53. 

§ 2.813 Written Communications 

This proposed rule would revise 

§ 2.813(a) to apply general requirements 

for correspondence with the 

Commission to communications 

concerning part 53, in addition to parts 

50, 52, and 100. 

§ 2.1103 Scope of Subpart K 

This proposed rule would revise the 

first sentence of § 2.1103 to extend the 

provisions of subpart K of part 2 to 

licenses under part 53 to expand the 

spent fuel capacity at the site of a 

civilian nuclear power plant. 

§ 2.1202 Authority and Role of NRC 

Staff 

This proposed rule would amend 

§ 2.1202 by revising paragraphs (a)(1) 

through (3), and (a)(6) to include 

references to part 53. 

§ 2.1301 Public Notice of Receipt of a 

License Transfer Application 

This proposed rule would revise 

§ 2.1301(b) to include a corresponding 

reference to license transfers under part 

53 in addition to parts 50 and 52. 

§ 2.1403 Authority and Role of the 

NRC Staff 

This proposed role would update 

§ 2.1403 to specify that ‘‘significant 

hazards considerations’’ has the same 

meaning as defined in part 53. 

§ 2.1500 Purpose and Scope 

This proposed rule would revise 

§ 2.1500 to extend the scope of subpart 

O of part 2 to DC rulemaking hearings 

under part 53. 

§ 2.1502 Commission Decision To 

Hold Legislative Hearing 

This proposed rule would revise 

§ 2.1502, paragraphs (a) and (b)(1) to 


lotter on DSK11XQN23PROD with PROPOSALS2
86990 Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules 

incorporate references to part 53 

regarding the Commission’s decision to 

hold a DC rulemaking. 

§ 10.1 Purpose 

This proposed rule would revise 

§ 10.1(a)(3) to include a reference to part 

53. 

§ 10.2 Scope 

This proposed rule would revise 

§ 10.2(b) to extend the scope of subpart 

A to applicants and holders of licenses, 

certificates, and standard design 

approvals under part 53 in addition to 

part 52. 

§ 11.7 Definitions 

This proposed rule would revise 

§ 11.7 such that terms defined in part 53 

have the same meaning when used in 

part 11. 

§ 19.2 Scope 

This proposed rule would revise 

§ 19.2(a) to include references to part 53. 

§ 19.3 Definitions 

This proposed rule would revise the 

definitions of ‘‘License’’ and ‘‘Regulated 

entities’’ in § 19.3 to incorporate 

references to part 53. 

§ 19.11 Posting of Notices to Workers 

This proposed rule would amend 

§ 19.11 by revising paragraphs (a), (b), 

and (e)(1) to apply to applicants and 

holders of licenses, permits, standard 

design approvals, and standard DCs 

under part 53 in addition to part 52. 

§ 19.14 Presence of Representatives of 

Licensees and Regulated Entities, and 

Workers During Inspections 

This proposed rule would revise 

§ 19.14(a) to apply to applicants and 

holders of a license, standard design 

approval, ESP, or standard DC under 

part 53 in addition to part 52. 

§ 19.20 Employee Protection 

This proposed rule would revise 

§ 19.20 to include a reference to 

protected activities under part 53. 

§ 20.1002 Scope 

This proposed rule would revise the 

first sentence of 10 CFR part 20, 

‘‘Standards for Protection Against 

Radiation,’’ § 20.1002 to extend the 

scope of part 20 to apply to persons 

licensed by the Commission to receive, 

use, transfer, or dispose of byproduct, 

source, or SNM or to operate a 

production or utilization facility under 

part 53. 

§ 20.1003 Definitions 

This proposed rule would revise 

§ 20.1003 to update the definition of 

‘‘License’’ to include those issued under 

part 53. 

§ 20.1101 Radiation Protection 

Programs 

This proposed rule would revise 

§ 20.1101(d) to exclude licensees subject 

to § 53.260 from its requirements. 

§ 20.1401 General Provisions and 

Scope 

This proposed rule would revise 

§ 20.1401, paragraphs (a) and (c) to 

extend the scope of subpart E of part 20 

to apply to the decommissioning of 

facilities licensed under part 53 and the 

release of part of a facility or site for 

unrestricted use in accordance with 

§ 53.1080. 

§ 20.1403 Criteria for License 

Termination Under Restricted 

Conditions 

This proposed rule would revise 

§ 20.1403(d) to include 

decommissioning plans under part 53. 

§ 20.1404 Alternate Criteria for License 

Termination 

This proposed rule would revise 

§ 20.1404(a)(4) to include a reference to 

part 53 regarding alternate criteria for 

license termination. 

§ 20.1406 Minimization of 

Contamination 

This proposed rule would revise 

§ 20.1406(a) to include references to 

applicants for licenses other than ESPs 

or MLs under part 53. It would also 

revise § 20.1406(b) to include references 

to standard DCs and standard design 

approvals under part 53 in addition to 

part 52. 

§ 20.1501 General 

This proposed rule would revise 

§ 20.1501(b) regarding the requirement 

for retention of records from surveys 

describing the location and amount of 

subsurface residual radioactivity at a 

site to include a reference to the 

retention requirements under part 53. 

§ 20.1905 Exemptions to Labeling 

Requirements 

This proposed rule would revise 

§ 20.1905(g) to apply to facilities 

licensed under part 53 in addition to 

parts 50 and 52 regarding exemptions to 

labeling requirements. 

§ 20.2004 Treatment or Disposal by 

Incineration 

This proposed rule would revise 

§ 20.2004(b)(1) to include references to 

part 53 regarding the treatment or 

disposal of waste oil by incineration. 

§ 20.2201 Reports of Theft or Loss of 

Licensed Material 

This proposed rule would revise 

§ 20.2201 to include references to part 

53 in paragraphs (a)(2)(i), (b)(2)(i) and 

(c) regarding requirements for reports of 

theft or loss of licensed material. 

§ 20.2202 Notification of Incidents 

This proposed rule would revise 

§ 20.2202(d)(1) to add references to part 

53 regarding reports to the NRC 

Operations Center. 

§ 20.2203 Reports of Exposures, 

Radiation Levels, and Concentrations of 

Radioactive Material Exceeding the 

Constraints or Limits 

This proposed rule would revise 

§ 20.2203(c) to refer to procedures under 

part 53 for reporting occurrences of 

exposures, radiation levels, and 

concentrations of radioactive material 

exceeding the constraints or limits. 

§ 20.2206 Reports of Individual 

Monitoring 

This proposed rule would revise 

§ 20.2206(a)(1) to include a reference to 

part 53. 

§ 21.2 Scope 

This proposed rule would revise 

§ 21.2, paragraphs (a), (b), and (c) to 

include references to part 53 regarding 

the scope and applicability of part 21 

requirements. 

§ 21.3 Definitions 

This proposed rule, in § 21.3 would 

revise the definitions of ‘‘Basic 

component,’’ ‘‘Commercial grade item,’’ 

‘‘Critical characteristics,’’ ‘‘Dedicating 

entity,’’ ‘‘Dedication,’’ ‘‘Defect,’’ and 

‘‘Substantial safety hazard’’ with 

references to part 53. 

§ 21.21 Notification of Failure To 

Comply or Existence of a Defect and Its 

Evaluation 

This proposed rule would revise 

§ 21.21, by incorporating references to 

part 53, to update the requirements for 

notifying the Commission of a failure to 

comply or defect in paragraphs (a)(3) 

and (d)(1). 

§ 21.51 Maintenance and Inspection of 

Records 

This proposed rule would revise 

§ 21.51(a)(4) and (5) to apply to 

applicants for standard DC and 

applicants or holders of a standard 

design approval under part 53, in 

addition to part 52, regarding the 

retention of records. 


lotter on DSK11XQN23PROD with PROPOSALS2
Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules 86991 

§ 21.61 Failure To Notify 

This proposed rule would revise 

§ 21.61(b) to include references to part 

53 licensees and applicants regarding 

failure to provide the notice required in 

§ 21.21. 

§ 25.5 Definitions 

This proposed rule would update the 

definition of ‘‘License’’ to include those 

issued under part 53. 

§ 25.17 Approval for Processing 

Applicants for Access Authorization 

This proposed rule would revise 

§ 25.17(a) to add a reference to part 53 

regarding AAs for individuals who need 

access to classified information in 

connection with activities under part 

53. 

§ 25.35 Classified Visits 

This proposed rule would update 

§ 25.35(a) to apply the requirements for 

classified visits to licensees, certificate 

holders, and applicants under part 53 in 

addition to part 52. 

§ 26.3 Scope 

This proposed rule would amend 

§ 26.3 by revising paragraph (d) and 

adding new paragraph (f) which would 

establish the phase of construction or 

operation by which applicants and 

licensees under part 53 would be 

required to comply with subpart M of 

part 26, or all of the requirements of part 

26 except subparts K and M. 

§ 26.4 FFD Program Applicability to 

Categories of Individuals 

This proposed rule would revise 

paragraphs (a), (b), (c), (e), (f), (g), and 

(h) of § 26.4 to include references to part 

53 and provisions for implementing an 

FFD program under subpart M. 

§ 26.5 Definitions 

This proposed rule would amend 

§ 26.5 by adding definitions for 

‘‘Biological marker,’’ ‘‘Change,’’ ‘‘Illicit 

substance,’’ ‘‘Reduction in FFD program 

effectiveness,’’ and ‘‘Special Nuclear 

Material.’’ It would also revise 

definitions of ‘‘Constructing or 

construction activities,’’ ‘‘Contractor/ 

vendor (C/V),’’ ‘‘Other entity,’’ 

‘‘Questionable validity,’’ ‘‘Reviewing 

official,’’ ‘‘Safety-related structures, 

systems, and components (SSCs),’’ 

‘‘Security-related SSCs,’’ and ‘‘Unit 

outage’’ within this section. 

§ 26.8 Information Collection 

Requirements: OMB Approval 

This proposed rule would revise 

§ 26.8(b) with the new information 

collection requirements contained in 

proposed §§ 26.202, 26.603, 26.604, 

26.605, 26.606, 26.607, 26.608, 26.609, 

26.611, 26.613, 26.617, and 26.619. 

§ 26.21 Fitness-for-Duty Program 

This proposed rule would revise 

§ 26.21 to include a reference to 

§ 26.3(f). 

§ 26.51 Applicability 

This proposed rule would revise 

§ 26.51 to extend the requirements of 

subpart C of part 26 to licensees and 

other entities identified in § 26.3(f) that 

do not implement the requirements of 

subpart M of part 26, as well as 

licensees and other entities that 

implement the requirements of § 26.605. 

§ 26.53 General Provisions 

This proposed rule would revise 

§ 26.53 paragraphs (e), (g), (h), and (i) to 

include references to § 26.3(f). 

§ 26.63 Suitable Inquiry 

This proposed rule would revise 

§ 26.63(d) with a reference to § 26.3(f). 

§ 26.73 Applicability 

This proposed rule would revise 

§ 26.73 to extend the requirements of 

subpart D of part 26 to licensees and 

other entities identified in § 26.3(f) that 

do not implement the requirements of 

subpart M of part 26, as well as 

licensees and other entities that 

implement the requirements of 

§ 26.605(b). 

§ 26.81 Purpose and Applicability 

This proposed rule would revise 

§ 26.81 to extend the requirements of 

subpart E of part 26 to licensees and 

other entities identified in § 26.3(f) that 

do not implement the requirements of 

subpart M of part 26, as well as 

licensees and other entities that 

implement the requirements of § 26.605. 

§ 26.201 Applicability 

This proposed rule would revise 

§ 26.201 to include references to the 

proposed provisions in §§ 26.3(f) and 

26.202, as well as revise the 

applicability of requirements in subpart 

I of part 26. 

§ 26.202 General Provisions for 

Facilities Licensed Under Part 53 

This proposed rule would add new 

§ 26.202, which would require 

applicable licensees under part 53 to 

incorporate a policy for fatigue 

management into their FFD program in 

accordance with the provisions of this 

section. 

§ 26.205 Work Hours 

This proposed rule would revise 

paragraphs (d)(7)(iii) and (d)(8) of 

§ 26.205 to incorporate references to 

§§ 26.606 and 26.202(a) and (b). 

§ 26.207 Waivers and Exceptions 

This proposed rule would revise 

§ 26.207(a)(1)(ii) to include references to 

§§ 26.608 and 26.202(c) and to include 

provisions for implementing certain 

face-to-face supervisor assessments 

using electronic communications. 

§ 26.211 Fatigue Assessments 

This proposed rule would revise 

§ 26.211, paragraphs (a)(1), (a)(3), and 

(b) to incorporate references to 

§§ 26.202(c), 26.607(b), 26.608, and 

26.619 and to include provisions for 

implementing certain face-to-face 

assessments using electronic 

communications. 

Subpart M—Fitness for Duty Programs 

for Facilities Licensed Under Part 53 

This proposed rule would add new 

Subpart M of part 26 containing 

§§ 26.601, 26.603, 26.604 through 

26.611, 26.613, 26.615, 26.617, and 

26.619, which adds an optional 

technology-inclusive, risk-informed, 

and performance-based approach for the 

application of drug and alcohol testing 

and fatigue management requirements 

for facilities licensed under part 53. 

§ 26.601 Applicability 

This proposed rule would add 

§ 26.601, which would allow a licensee 

or other entity in § 26.3(f) to establish an 

FFD program in accordance with the 

requirements of subpart M of part 26. 

§ 26.603 General Provisions 

This proposed rule would add 

§ 26.603, which would establish the 

general requirements for implementing 

an FFD program under subpart M of part 

26. 

§ 26.604 FFD Program Requirements 

for Facilities That Satisfy the § 26.603(c) 

Criterion 

This proposed rule would add 

§ 26.604, which would establish the 

FFD program elements for a licensee or 

other entity whose facilities and 

operations demonstrate compliance 

with the criterion in § 26.603(c). 

§ 26.605 FFD Program Requirements 

for Facilities That Do Not Implement 

§ 26.604 

This proposed rule would add 

§ 26.605, which would establish the 

FFD program elements for a licensee or 

other entity that does not demonstrate 

compliance with the criterion in 

§ 26.603(c), or otherwise chooses to 

maintain an FFD program under this 

section. 


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§ 26.606 Written Policies and 

Procedures 

This proposed rule would add 

§ 26.606, which would require licensees 

and other entities that implement an 

FFD program under subpart M of part 26 

to develop a written FFD policy 

statement and provide it to all 

individuals subject to the FFD program, 

and to establish, implement, and 

maintain written procedures addressing 

the topics outlined in this section. 

§ 26.607 Drug and Alcohol Testing 

This proposed rule would add 

§ 26.607, which would establish 

requirements for licensees and other 

entities performing drug and alcohol 

testing as part of an FFD program under 

subpart M of part 26. 

§ 26.608 FFD Program Training 

This proposed rule would add 

§ 26.608, which would require 

individuals who are subject to the FFD 

program under subpart M of part 26 to 

receive periodic training on FFD 

policies and procedures, including their 

duties and responsibilities under the 

BOP. 

§ 26.609 Behavioral Observation 

This proposed rule would add 

§ 26.609, which would establish the 

requirements for a BOP under subpart M 

of part 26. 

§ 26.610 Sanctions 

This proposed rule would add 

§ 26.610, which would require licensees 

and other entities implementing an FFD 

program under subpart M of part 26 to 

establish sanctions for FFD policy 

violations. 

§ 26.611 Protection of Information 

This proposed rule would add 

§ 26.611, which would require licensees 

and other entities implementing an FFD 

program under subpart M of part 26 to 

establish a system to protect personal 

information against unauthorized 

disclosure. 

§ 26.613 Appeals Process 

This proposed rule would add 

§ 26.613, which would require licensees 

and other entities that implement an 

FFD program under subpart M of part 26 

to establish procedures for an individual 

to appeal a policy violation 

determination. 

§ 26.615 Audits 

This proposed rule would add 

§ 26.615, which would establish 

provisions for licensees and other 

entities that implement an FFD program 

under subpart M of part 26 to conduct 

audits to monitor the effectiveness of 

FFD program elements. 

§ 26.617 Recordkeeping and Reporting 

This proposed rule would add 

§ 26.617, which would require licensees 

or other entities implementing an FFD 

program under subpart M of part 26 to 

retain records pertaining to the 

administration of the program and to 

make reports in accordance with the 

requirements of this section. 

§ 26.619 Suitability and Fitness 

Determinations 

This proposed rule would add 

§ 26.619, which would require licensees 

and other entities that implement FFD 

programs to develop, implement, and 

maintain procedures to assess whether 

individuals are fit to perform the duties 

that make them subject to the FFD 

program. 

§ 26.709 Applicability 

This proposed rule would designate 

the current paragraph as new paragraph 

(a), and it would be revised to reference 

paragraphs (a) through (d) of § 26.3. It 

would also add paragraph (b) to 

§ 26.709, which would extend the 

requirements of subpart N of part 26 to 

licensees and other entities identified in 

§ 26.3(f) that do not implement the 

requirements of subpart M of part 26, as 

well as licensees and other entities that 

implement the requirements of 

§ 26.605(b). 

§ 26.711 General Provisions 

This proposed rule would revise 

§ 26.711(c) and (d) to incorporate a 

reference to § 26.3(f). 

§ 26.825 Criminal Penalties 

This proposed rule would revise 

§ 26.825(b) to include a reference to the 

proposed § 26.601. 

§ 30.4 Definitions 

This proposed rule would revise the 

definition for ‘‘Utilization facility’’ in 

§ 30.4 to include utilization facilities 

defined in the regulations under part 53 

in addition to part 50. 

§ 30.50 Reporting Requirements 

This proposed rule would revise 

§ 30.50(c)(3) to include references to 

part 53 in addition to part 50. 

§ 40.60 Reporting Requirements 

This proposed rule would revise 

§ 40.60(c)(3) to include references to 

part 53 in addition to part 50 regarding 

reporting requirements. 

§ 50.47 Emergency Plans 

This proposed rule would revise 

§ 50.47(a)(1) and (e) with appropriate 

references to part 53. 

§ 50.54 Conditions of Licenses 

This proposed rule would revise 

§ 50.54(q)(2), (q)(4), and (gg)(1) with 

appropriate references to part 53. 

§ 50.160 Emergency Preparedness for 

Small Modular Reactors, Non-LightWater Reactors, and Non-Power 

Production or Utilization Facilities 

This proposed rule would revise 

§ 50.160(b)(3) and (c)(2) with the 

appropriate references to part 53. 

Appendix B to 10 CFR Part 50—Quality 

Assurance Criteria for Nuclear Power 

Plants and Fuel Reprocessing Plants 

This proposed rule would revise 

appendix B to part 50 by revising the 

introduction and specific criteria to 

incorporate the appropriate references 

and terminology for part 53. 

§ 51.20 Criteria for and Identification 

of Licensing and Regulatory Actions 

Requiring Environmental Impact 

Statements 

This proposed rule would revise 

§ 51.20(b)(1) and (2) to require an EIS 

prior to the issuance of a CP, LWA, or 

ESP under part 53, or the issuance to 

renewal of a full power or design 

capacity license to operate a nuclear 

power reactor, testing facility, or fuel 

reprocessing plant under part 53. 

§ 51.22 Criterion for Categorical 

Exclusion; Identification of Licensing 

and Regulatory Actions Eligible for 

Categorical Exclusion or Otherwise Not 

Requiring Environmental Review 

This proposed rule would revise 

§ 51.22 to include corresponding 

references to part 53 in paragraphs 

(c)(3), (c)(9), (c)(12), (c)(17), (c)(22) and 

(23). 

§ 51.26 Requirement To Publish Notice 

of Intent and Conduct Scoping Process 

This proposed rule would revise 

§ 51.26(d) to add a reference to part 53. 

§ 51.30 Environmental Assessment 

This proposed rule would revise the 

introductory text to paragraph (a) and 

revise paragraphs (d) and (e) of § 51.30 

to incorporate the appropriate 

references to part 53 regarding EAs. 

§ 51.31 Determinations Based on 

Environmental Assessment 

This proposed rule would revise 

§ 51.31(a) to include a reference to part 

53. 


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§ 51.32 Finding of No Significant 

Impact 

This proposed rule would revise 

§ 51.32(b)(1) and (3), finding there is no 

significant environmental impact 

associated with the issuance of standard 

DCs and MLs under part 53. 

§ 51.49 Environmental Report-Limited 

Work Authorization 

This proposed rule would revise the 

introductory text of § 51.49(c) to require 

applicants for an ESP under part 53 

requesting a LWA to include the 

environmental report required by 

§ 51.50(b). 

§ 51.50 Environmental Report— 

Construction Permit, Early Site Permit, 

or Combined License Stage 

This proposed rule would revise 

§ 51.50, paragraphs (a), (b)(4), and the 

introductory text for paragraph (c) to 

incorporate the appropriate references 

to part 53. 

§ 51.53 Postconstruction 

Environmental Reports 

This proposed rule would revise 

§ 51.53(d) to include the appropriate 

references to part 53 regarding a license 

termination plan or decommissioning 

plan and related requirements for 

postconstruction environmental reports. 

§ 51.54 Environmental Report— 

Manufacturing License 

This proposed rule would update 

§ 51.54(a) to require applicants for MLs 

under part 53 to submit an 

environmental report with the 

application. 

§ 51.55 Environmental Report— 

Standard Design Certification 

This proposed rule would update 

§ 51.55(a) to require applicants for a 

standard DC under part 53 to submit an 

environmental report with the 

application. 

§ 51.58 Environmental Report— 

Number of Copies; Distribution 

This proposed rule would revise 

§ 51.58(b) to incorporate the appropriate 

references to part 53. 

§ 51.77 Distribution of Draft 

Environmental Impact Statement 

This proposed rule would revise the 

introductory text for § 51.77(a) to add a 

reference to part 53. 

§ 51.92 Supplement to the Final 

Environmental Impact Statement 

This proposed rule would revise 

§ 51.92(b) to apply to COL applications 

referencing an ESP under part 53. 

§ 51.95 Postconstruction 

Environmental Impact Statements 

This proposed rule would revise the 

introductory text for § 51.95(c) to 

include a reference to part 53 regarding 

the Commission’s obligations to prepare 

an EIS following the renewal of an 

operating or COL for a nuclear plant 

under part 53. 

§ 51.101 Limitations on Actions 

This proposed rule would revise 

§ 51.101(a)(2) to include the 

corresponding references to part 53 

where appropriate. 

§ 51.103 Record of Decision—General 

This proposed rule would update 

§ 51.103(a)(6) to apply to the issuance of 

a LWA in connection with a CP or COL 

under part 53. 

§ 51.105 Public Hearings in 

Proceedings for Issuance of 

Construction Permits or Early Site 

Permits; Limited Work Authorizations 

This proposed rule would revise 

§ 51.105(c)(1) to include the appropriate 

reference to LWAs under part 53 for CPs 

or ESPs. 

§ 51.107 Public Hearings in 

Proceedings for Issuance of Combined 

Licenses; Limited Work Authorizations 

This proposed rule would amend 

§ 51.107 by revising the introductory 

text for paragraphs (a) and (b) and 

updating paragraph (d)(1) to include the 

appropriate corresponding references to 

part 53. 

§ 51.108 Public Hearings on 

Commission Findings That Inspections, 

Tests, Analyses, and Acceptance 

Criteria of Combined Licenses Are Met 

This proposed rule would revise 

§ 51.108 to incorporate the appropriate 

references to part 53. 

10 CFR part 53—Risk-Informed, 

Technology-Inclusive Regulatory 

Framework for Commercial Nuclear 

Plants 

This proposed rule would add a new 

part to 10 CFR Chapter I, designated as 

Part 53 including §§ 53.000 through 

53.9010. 

§ 53.000 Purpose 

This proposed rule would add 

§ 53.000 which provides an optional 

technology-inclusive, performancebased framework for the issuance, 

amendment, renewal, and termination 

of licenses, permits, certifications, and 

approvals for commercial nuclear plants 

licensed under section 103 of the 

Atomic Energy Act of 1954, as amended. 

Subpart A—General Provisions 

This proposed rule would add subpart 

A, to establish a set of general 

provisions, which apply to all 

applicants and licensees under part 53. 

§ 53.015 Scope 

This proposed rule would add 

§ 53.015, which would extend the 

provisions of subpart A to all applicants 

and licensees under part 53. 

§ 53.020 Definitions 

This proposed rule would add 

§ 53.020, which would define key terms 

in part 53. 

§ 53.040 Written Communications 

This proposed rule would add 

§ 53.040, which would govern how 

applicants and licensees submit written 

communications to the NRC, including 

applications, submissions related to the 

security plans, emergency plan, and 

quality assurance, certifications of 

permanent cessation of operations and 

permanent fuel removal, and other 

submittals required under part 53. 

§ 53.050 Deliberate Misconduct 

This proposed rule would add 

§ 53.050, which would prohibit 

licensees or applicants, contractors and 

subcontractors, or employees of those 

entities from deliberately violating NRC 

rules, regulations, or orders, or the 

terms, conditions, and limitations of a 

part 53 license. This proposed rule 

would also prohibit deliberate 

submissions of incomplete or inaccurate 

information. Violations would be 

subject to enforcement actions under 

subpart B of part 2. 

§ 53.060 Employee Protection 

This proposed rule would add 

§ 53.060, which would prohibit 

applicants and licensees from 

discriminating against employees for 

engaging in the protected activities 

listed in this section and provide 

remedial procedures for employees who 

believe they are the subjects of 

discrimination. 

§ 53.070 Completeness and Accuracy 

of Information 

This proposed rule would add 

§ 53.070, which would require licensees 

and applicants under part 53 to provide 

complete and accurate information in 

accordance with all applicable laws, 

Commission regulations, and the terms 

and conditions of their license. This 

proposed rule would also require 

licensees to notify the Commission 

within two days of identifying 

information with material implications 


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for public health and safety or common 

defense and security. 

§ 53.080 Specific Exemptions 

This proposed rule would add 

§ 53.080, which would establish the 

special circumstances under which the 

Commission could grant exemptions to 

part 53 licensees and the Commission’s 

criteria for making such a 

determination. 

§ 53.090 Standards for Review 

This proposed rule would add 

§ 53.090 to establish the standards that 

the Commission would consider when 

determining whether to issue a permit 

or license under part 53. 

§ 53.100 Jurisdictional Limits 

This proposed rule would add 

§ 53.100, which would provide that 

permits, licenses, standard design 

approvals, and standard DCs are solely 

issued for activities within the 

jurisdiction of the United States. 

§ 53.110 Attacks and Destructive Acts 

This proposed rule would add 

§ 53.110, which would exempt licensees 

or applicants under part 53 from 

providing design features to protect 

against attacks or destructive acts 

directed at the facility by United States 

adversaries. 

§ 53.115 Rights Related to Special 

Nuclear Material 

This proposed rule would add 

§ 53.115, which would establish 

provisions regarding the rights to SNM 

under a part 53 license. 

§ 53.117 License Suspension and 

Rights of Recapture 

This proposed rule would add 

§ 53.117, which would provide that the 

Commission may suspend licenses and 

recapture material or control of a facility 

in a state of war or national emergency 

declared by Congress. 

§ 53.120 Information Collection 

Requirements: OMB Approval 

This proposed rule would add 

§ 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 technologyinclusive performance standards that 

would be used throughout part 53 to 

determine appropriate regulatory 

controls for SSCs, human actions, and 

programs. 

§ 53.210 Safety Criteria for DesignBasis Accidents 

This proposed rule would add 

§ 53.210 to set dose values to ensure that 

plants are designed to limit the public’s 

radiation exposure in the event of a 

DBA. 

§ 53.220 Safety Criteria for LicensingBasis Events Other Than Design-Basis 

Accidents 

This proposed rule would add 

§ 53.220 to require plants to implement 

a combination of design features and 

programmatic controls to control risks 

to the public in the event of a LBE other 

than a DBA. 

§ 53.230 Safety Functions 

This proposed rule would add 

§ 53.230, which specifies that limiting 

the release of radioactive materials from 

the facility is the primary safety 

function of a commercial nuclear plant, 

and that additional safety functions 

must be defined to support the retention 

of radioactive materials during LBEs. 

§ 53.240 Licensing-Basis Events 

This proposed rule would add 

§ 53.240 to require commercial nuclear 

plants to conduct an analysis of LBEs to 

confirm that design features and 

programmatic controls satisfy the safety 

criteria under §§ 53.210 and 53.220, or 

alternatively, under § 53.470. 

§ 53.250 Defense in Depth 

This proposed rule would add 

§ 53.250 to establish a performancebased, defense-in-depth approach to 

address uncertainties about the 

effectiveness and reliability of plant 

SSCs, personnel, and programmatic 

controls. 

§ 53.260 Normal Operations 

This proposed rule would add 

§ 53.260, requiring holders of licenses to 

operate commercial nuclear plants to 

control public doses and dose rates in 

unrestricted areas to meet the 

requirements in part 20, during normal 

plant operation. 

§ 53.270 Protection of Plant Workers 

This proposed rule would add 

§ 53.270, requiring holders of licenses to 

operate commercial nuclear plants to 

control occupational doses to meet the 

requirements in part 20. 

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. 

§ 53.400 Design Features for LicensingBasis Events 

This proposed rule would add 

§ 53.400, which would require design 

features that satisfy the safety criteria 

defined in § 53.210 and § 53.220 or 

§ 53.470 and fulfill the safety functions 

identified in § 53.230 during LBEs. 

§ 53.410 Functional Design Criteria for 

Design-Basis Accidents 

This proposed rule would add 

§ 53.410, which would stipulate that 

functional design criteria must be 

defined for each design feature required 

by § 53.400 to demonstrate compliance 

with the safety criteria defined in 

§ 53.210 for DBAs. 

§ 53.415 Protection Against External 

Hazards 

This proposed rule would add 

§ 53.415, which would require SR SSCs 

to be designed to withstand the effects 

of natural phenomena and constructed 

hazards while performing the intended 

safety functions. 

§ 53.420 Functional Design Criteria for 

Licensing-Basis Events Other Than 

Design-Basis Accidents 

This proposed rule would add 

§ 53.420, which would require 

functional design criteria to be defined 

for each design feature required by 

§ 53.400 to demonstrate compliance 

with the safety criteria defined in 

§ 53.220 for LBEs other than DBAs. 

§ 53.425 Design Features and 

Functional Design Criteria for Normal 

Operations 

This proposed rule would add 

§ 53.425, which would require 

commercial nuclear plants to implement 

design features and define functional 

design criteria sufficient to demonstrate 

compliance with § 53.850 and show 

through functional design criteria that 

design features and corresponding 

programmatic controls control wastes, 

as required under part 20. 

§ 53.430 Design Features and 

Functional Design Criteria for Protection 

of Plant Workers 

This proposed rule would add 

§ 53.430, which would require 

commercial nuclear plants to implement 

design features and define functional 

design criteria sufficient to demonstrate 

compliance with § 53.270. 

§ 53.440 Design Requirements 

This proposed rule would add 

§ 53.440, which would establish various 


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design feature requirements, including 

protection against fires and explosions, 

criticality accidents, and the impact of 

a large commercial aircraft. 

§ 53.450 Analysis Requirements 

This proposed rule would add 

§ 53.450, which would require 

commercial nuclear plants to perform 

PRAs in combination with other 

analytical methods to identify and 

assess risks and determine compliance 

with the safety criteria in subpart B. In 

addition, § 53.450 would require 

analysis of DBAs and other analyses to 

assess the adequacy of protections 

against fire, aircraft impact, and the 

release of effluents. 

§ 53.460 Safety Categorization and 

Special Treatments 

This proposed rule would add 

§ 53.460 to address the safety 

classification of SSCs and determine 

appropriate special treatments. 

§ 53.470 Maintaining Analytical Safety 

Margins Used To Justify Operational 

Flexibilities 

This proposed rule would add 

§ 53.470 to permit applicants and 

licensees to implement more restrictive 

criteria than that defined in §§ 53.220 

and 53.450(e) to support operational 

flexibilities. 

§ 53.480 Earthquake Engineering 

This proposed rule would add 

§ 53.480 to provide overall seismic 

design considerations based on the 

safety criteria in subpart B and siting 

requirements in subpart D to ensure that 

SSCs are able to withstand the effects of 

earthquakes without loss of capability to 

fulfill safety functions. 

Subpart D—Siting Requirements 

This proposed rule would add subpart 

D, which would address requirements 

associated with the siting of commercial 

nuclear facilities under part 53, 

including considerations of external 

hazards and potential adverse impacts 

on the surrounding population. 

§ 53.500 General Siting and Siting 

Assessment 

This proposed rule would add 

§ 53.500, which would require a siting 

assessment for each commercial nuclear 

plant to ensure that design features and 

programmatic controls are sufficient to 

address LBEs and mitigate potential 

adverse impacts of the plant on the 

surrounding environs. 

§ 53.510 External Hazards 

This proposed rule would add 

§ 53.510, which would require sitespecific 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. 

§ 53.520 Site Characteristics 

This proposed rule would add 

§ 53.520, which would require the 

design and analyses conducted under 

subpart C to consider how site 

characteristics may contribute to LBEs. 

§ 53.530 Population-Related 

Considerations 

This proposed rule would add 

§ 53.530, which would establish 

requirements related to the facility’s 

exclusion area, low-population zone, 

and population center distance. 

§ 53.540 Siting Interfaces 

This proposed rule would add 

§ 53.540, which would require that 

external hazards and site characteristics 

must be accounted for in the design 

features, programmatic controls, and 

supporting analyses used to 

demonstrate compliance with the safety 

criteria in §§ 53.210 and 53.220. 

Subpart E—Construction and 

Manufacturing Requirements 

This proposed rule would add subpart 

E, which would establish requirements 

for the construction and manufacture of 

commercial nuclear plants. 

§ 53.600 Construction and 

Manufacturing—Scope and Purpose 

This proposed rule would add 

§ 53.600, which would indicate that this 

subpart applies to construction and 

manufacturing activities authorized by a 

CP, COL, ML, or LWA issued under this 

part. 

§ 53.605 Reporting of Defects and 

Noncompliance 

This proposed rule would add 

§ 53.605, which would describe the 

procedures, notification requirements, 

and records retention requirements that 

each CP, ML, and COL is subject to with 

respect to reporting of defects and 

noncompliance. 

§ 53.610 Construction 

This proposed rule adds § 53.610 to 

address the management and control of 

the construction of a commercial 

nuclear plant, including specific 

requirements for procedures and quality 

assurance, control of radioactive 

materials, and post construction 

inspections. 

§ 53.620 Manufacturing 

This proposed rule would add 

§ 53.620, which would ensure that the 

holders of an ML under part 53 develop 

plans, programs, and organizational 

units to manage and control 

manufacturing activities, and would 

establish requirements for the loading of 

fuel into a manufactured reactor for 

subsequent transport to a commercial 

nuclear plant and operation pursuant to 

a COL. 

Subpart F—Requirements for Operation 

This proposed rule would add subpart 

F, which would establish regulatory 

requirements to ensure that the safety 

criteria in subpart B are satisfied 

whenever a commercial nuclear plant 

licensed under part 53 is operational. 

This includes periods of normal 

operation and unplanned events. 

§ 53.700 Operational Objectives 

This proposed rule would add 

§ 53.700, which would establish general 

operational objectives to ensure that 

licensees under part 53 have 

implemented and maintained the SSCs 

necessary to demonstrate compliance 

with the safety functions identified in 

subpart B for addressing normal 

operations and responding to LBEs. 

§ 53.710 Maintaining Capabilities and 

Availability of Structures, Systems, and 

Components 

This proposed rule would add 

§ 53.710, which would require licensees 

under part 53 to demonstrate 

compliance with the safety criteria in 

subpart B by establishing TS for all SR 

SSCs and developing documents and 

procedures for all NSRSS SSCs. 

§ 53.715 Maintenance, Repair, and 

Inspection Programs 

This proposed rule would add 

§ 53.715, which would require licensees 

to develop, implement, and maintain 

programs to assess and manage any risks 

posed by maintenance activities and to 

evaluate the efficacy of performance, 

condition monitoring, and maintenance 

activities. 

§ 53.720 Response to Seismic Events 

This proposed rule would add 

§ 53.720, which would establish 

requirements for licensees to respond to 

a seismic event during the operating 

phase of the life cycle of a commercial 

nuclear plant. 

§ 53.725 General Staffing, Training, 

Personnel Qualifications, and Human 

Factors Requirements 

This proposed rule would add 

§ 53.725, which would provide an 


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overview of the staffing, training, 

personnel qualifications, and human 

factors requirements established in 

§§ 53.725 through 53.830 and would 

provide definitions of ‘‘Automation,’’ 

‘‘Auxiliary operator,’’ ‘‘Controls,’’ 

‘‘Generally licensed reactor operator,’’ 

‘‘Load following,’’ ‘‘Operator,’’ 

‘‘Performance testing,’’ ‘‘Reference 

plant,’’ ‘‘Self-reliant mitigation facility,’’ 

‘‘Senior operator,’’ ‘‘Simulation 

facility,’’ and ‘‘Systems approach to 

training.’’ Proposed §§ 53.725 through 

53.830 would apply to applicants for or 

holders of OLs or COLs under part 53. 

§ 53.726 Communications 

This proposed rule would add 

§ 53.726, which would contain 

communications requirements 

applicable to sections §§ 53.725 through 

53.830. It also contains requirements to 

notify the Commission within 30 days 

should a specifically licensed operator 

or senior operator be reassigned, 

terminated, or suffer permanent 

disability or illness. 

§ 53.728 Completeness and Accuracy 

of Information 

This proposed rule would add 

§ 53.728, which would require 

submitted information to be complete 

and accurate in all material respects. 

§ 53.730 Defining, Fulfilling, and 

Maintaining the Role of Personnel in 

Ensuring Safe Operations 

This proposed rule would add 

§ 53.730, which would establish 

technical requirements for applicants or 

holders of OLs or COLs within the areas 

of HFE, human-system interface design, 

concept of operations, functional 

requirements analysis, function 

allocation, operating experience, 

procedures, staffing, operator training, 

operator examinations, and operator 

proficiency. 

§ 53.735 General Exemptions 

This proposed rule would add 

§ 53.735, which would establish general 

exemptions for licensed operators. 

§ 53.740 Facility Licensee 

Requirements—General 

This proposed rule would add 

§ 53.740, which would establish staffing 

requirements for interaction-dependentmitigation facilities and self-reliant 

mitigation facilities. 

§ 53.745 Operator License 

Requirements 

This proposed rule would add 

§ 53.745, which would require 

individuals to be licensed to perform 

certain functions. 

§ 53.760 Operator Licensing 

This proposed rule would add 

§ 53.760, which would address the 

applicability of the requirements of 

§§ 53.760 through 53.795 for specifically 

licensed operators and senior operators. 

§ 53.765 Medical Requirements 

This proposed rule would add 

§ 53.765, which would establish 

medical requirements for specifically 

licensed operators and senior operators. 

§ 53.770 Incapacitation Because of 

Disability or Illness 

This proposed rule would add 

§ 53.770, which would establish 

requirements to address permanent 

medical conditions for specifically 

licensed operators and senior operators. 

§ 53.775 Applications for Operators 

and Senior Operators 

This proposed rule would add 

§ 53.775, which would establish the 

application process and requirements 

for individuals applying for specific 

operator and senior operator licenses. 

§ 53.780 Training, Examination, and 

Proficiency Program 

This proposed rule would add 

§ 53.780, which would contain the 

requirements associated with 

specifically licensed operator and senior 

operator initial training, initial 

examinations, requalification training, 

requalification examinations, 

examination integrity, simulation 

facilities, waivers, and proficiency. 

§ 53.785 Conditions of Operator and 

Senior Operator Licenses 

This proposed rule would add 

§ 53.785, which would establish 

conditions for specific operator and 

senior operator licenses. 

§ 53.790 Issuance, Modification, and 

Revocation of Operator and Senior 

Operator Licenses 

This proposed rule would add 

§ 53.790, which would contain 

requirements associated with the 

issuance, modification, or revocation of 

specific operator and senior operator 

licenses. 

§ 53.795 Expiration and Renewal of 

Operator and Senior Operator Licenses 

This proposed rule would add 

§ 53.795, which would contain 

requirements associated with the 

expiration and renewal of specific 

operator and senior operator licenses. 

§ 53.800 Facility Licensees for SelfReliant-Mitigation Facilities 

This proposed rule would add 

§ 53.800, which would establish the 

technical criteria by which commercial 

nuclear plants under part 53 are 

determined to be of the self-reliant 

mitigation class of facilities that would 

be staffed by GLROs in lieu of 

specifically licensed operators and 

senior operators. 

§ 53.805 Facility Licensee 

Requirements Related to Generally 

Licensed Reactor Operators 

This proposed rule would add 

§ 53.805, which would establish 

requirements that apply to the facility 

licensee at those facilities staffed by 

GLROs. 

§ 53.810 Generally Licensed Reactor 

Operators 

This proposed rule would add 

§ 53.810, which would issue and 

describe the general license for GLROs 

that manipulate the controls of a selfreliant mitigation facility. 

§ 53.815 Generally Licensed Reactor 

Operator Training, Examination, and 

Proficiency Programs 

This proposed rule would add 

§ 53.815, which would contain the 

requirements for GLRO initial training, 

initial examinations, continuing 

training, requalification examinations, 

examination integrity, simulation 

facilities, examination waivers, and 

proficiency. 

§ 53.820 Cessation of Individual 

Applicability 

This proposed rule would add 

§ 53.820, which would address the 

requirements by which the general 

license for GLROs would cease to be 

applicable on an individual basis. 

§ 53.830 Training and Qualification of 

Commercial Nuclear Plant Personnel 

This proposed rule would add 

§ 53.830, which would address training 

and qualification requirements for 

supervisors, technicians, and other 

appropriate operating personnel at 

commercial nuclear plants. 

§ 53.845 Programs 

This proposed rule would add 

§ 53.845, which would require licensees 

under part 53 to establish programs that 

include, but are not limited to, radiation 

protection, emergency preparedness, 

security, quality assurance, integrity 

assessment, fire protection, ISI and IST, 

and facility safety, to ensure that the 

safety criteria and functions in subpart 


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B are maintained during normal 

operations and LBEs. 

§ 53.850 Radiation Protection 

This proposed rule would add 

§ 53.850, which would require licensees 

under part 53 to implement and 

maintain programs and processes to 

limit and monitor radioactive plant 

effluents and limit the exposure of plant 

personnel and the public. 

§ 53.855 Emergency Preparedness 

This proposed rule would add 

§ 53.855, which would require licensees 

under this part to have an emergency 

response plan for radiological 

emergencies. 

§ 53.860 Security Programs 

This proposed rule would add 

§ 53.860, which would require licensees 

under part 53 to develop, implement, 

and maintain programs for physical 

security, FFD, AA, cybersecurity, and 

information security. 

§ 53.865 Quality Assurance 

This proposed rule would add 

§ 53.865, which would require licensees 

under part 53 to establish a quality 

assurance program that includes a 

written manual to ensure activities are 

conducted in accordance with codes 

and standards found acceptable by the 

NRC. 

§ 53.870 Integrity Assessment 

Programs 

This proposed rule would add 

§ 53.870, which would require licensees 

under part 53 to establish an integrity 

assessment program to ensure that the 

plant continues to fulfill safety criteria 

and functional design criteria as it ages. 

§ 53.875 Fire Protection 

This proposed rule would add 

§ 53.875, which would require licensees 

under part 53 to establish a fire 

protection plan and describe the 

necessary elements that the plan must 

incorporate. 

§ 53.880 Inservice Inspection and 

Inservice Testing 

This proposed rule would add 

§ 53.880, which would require licensees 

under part 53 to develop and implement 

a program for ISI and IST in accordance 

with the requirements of this section. 

§ 53.910 Procedures and Guidelines 

This proposed rule would add 

§ 53.910, which would require licensees 

under part 53 to develop, maintain, and 

implement procedures and guidelines 

that address normal plant operations 

and responses to unplanned events. 

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. 

§ 53.1000 Scope and Purpose 

This proposed rule would add 

§ 53.1000, which would establish the 

scope of the decommissioning 

requirements for applicants and 

licensees under part 53 and describe the 

contents of subpart G of part 53. 

§ 53.1010 Financial Assurance for 

Decommissioning 

This proposed rule would add 

§ 53.1010, which would establish the 

requirement that applicants for an OL or 

COL under part 53 provide reasonable 

assurance that funds will be available 

for the decommissioning process. This 

section would describe the requirements 

associated with the required plan and 

an associated decommissioning report 

that ensures and documents that 

adequate funding for decommissioning 

will be available. 

§ 53.1020 Cost Estimates for 

Decommissioning 

This proposed rule would add 

§ 53.1020, which would require sitespecific cost estimates for 

decommissioning and establish the 

aspects that must be included in the 

estimate. 

§ 53.1030 Annual Adjustments to Cost 

Estimates for Decommissioning 

This proposed rule would add 

§ 53.1030, which would require that 

holders of an OL or COL under part 53 

annually adjust their cost estimate for 

decommissioning to account for 

escalation in labor, energy, and waste 

burial costs. This section would allow 

licensees to elect either a site-specific 

adjustment factor or a generic 

adjustment factor. 

§ 53.1040 Methods for Providing 

Financial Assurance for 

Decommissioning 

This proposed rule would add 

§ 53.1040, which would establish 

suitable methods that holders of an OL 

or COL under part 53 may use to 

provide financial assurance for 

decommissioning to the NRC. 

§ 53.1045 Limitations on the Use of 

Decommissioning Trust Funds 

This proposed rule would add 

§ 53.1045, which would establish 

requirements for decommissioning trust 

funds under part 53, including criteria 

for using decommissioning trust funds 

and required terms. 

§ 53.1050 NRC Oversight 

This proposed rule would add 

§ 53.1050, which would outline the 

steps the NRC may take to ensure 

adequate accumulation of 

decommissioning funds. 

§ 53.1060 Reporting and 

Recordkeeping Requirements 

This proposed rule would add 

§ 53.1060, which would contain 

reporting and recordkeeping 

requirements related to 

decommissioning for each holder of an 

OL or COL under part 53. This section 

would outline requirements for 

documents such as: certification of 

decommissioning funding, 

decommissioning cost estimates and 

copies of financial instruments, licensee 

records of information important to safe 

and effective decommissioning, postshutdown decommissioning activities 

report, financial assurance reports, and 

reports on the status of funding for 

managing irradiated fuel. 

§ 53.1070 Termination of License 

This proposed rule would add 

§ 53.1070, which would establish 

procedures for decommissioning and 

license termination applicable to 

licensees under part 53 that have 

determined to permanently cease 

operations. 

§ 53.1075 Program Requirements 

During Decommissioning 

This proposed rule would add 

§ 53.1075, which would require 

licensees under part 53 to establish and 

maintain a decommissioning fire 

protection program to prevent, detect, 

and control fires, and ensure that the 

risk of fire induced radiological hazards 

are minimized through the various 

stages of facility decommissioning. 

§ 53.1080 Release of Part of a 

Commercial Nuclear Plant or Site for 

Unrestricted Use 

This proposed rule would add 

§ 53.1080, which would establish 

licensee procedures for requesting and 

NRC procedures for approving partial 

release of a commercial nuclear plant or 

site for unrestricted use prior to 

receiving approval of a license 

termination plan from the Commission 

under part 53. 

Subpart H—Licenses, Certifications, and 

Approvals 

This proposed rule would add subpart 

H, which would govern the process of 

applying for, amending, renewing, or 


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terminating a LWA, ESP, standard 

design approval, standard DC, ML, CP, 

OL, or COL under part 53. 

§ 53.1100 Filling of Application for 

Licenses, Certifications, or Approvals; 

Oath or Affirmation 

This proposed rule would add 

§ 53.1100, which would establish 

requirements for applicants seeking a 

standard design approval, standard DC, 

license, or permit under part 53 to 

submit an application. 

§ 53.1101 Requirement for License 

This proposed rule would add 

§ 53.1101, which would prohibit any 

use of a utilization facility except as 

authorized by a license issued by the 

NRC or by an exception as described in 

§ 53.1120. 

§ 53.1103 Combining Applications and 

Licenses 

This proposed rule would add 

§ 53.1103, which would permit 

applicants under part 53 seeking 

multiple licenses to submit a single 

application, and the Commission to 

issue a single license for activities that 

would otherwise be licensed separately. 

§ 53.1106 Elimination of Repetition 

This proposed rule would add 

§ 53.1106, which would allow 

applicants under part 53 to reference 

information contained in previous 

documents filed with the Commission 

so long as those references are clear and 

specific. 

§ 53.1109 Contents of Applications; 

General Information 

This proposed rule would add 

§ 53.1109, which would establish the 

general content to be included in 

applications made under part 53, 

including but not limited to the 

identifying information of the applicant 

and the radiological emergency 

response plans of government entities 

within the plume exposure pathway 

EPZ. 

§ 53.1112 Environmental Conditions 

This proposed rule would add 

§ 53.1112, which would allow the 

Commission to attach conditions to CPs, 

ESPs, and licenses issued under part 53 

to address environmental issues during 

construction, operation, or 

decommissioning. These conditions will 

be derived from the information 

contained in the environmental report 

submitted as part of the application for 

a permit or license. 

§ 53.1115 Agreement Limiting Access 

to Classified Information 

This proposed rule would add 

§ 53.1115, which would require 

applicants to agree in writing, prior to 

receiving a license or standard design 

approval under part 53, to restrict any 

facilities, or any individuals with access 

to plant facilities, from possessing 

Restricted Data or classified National 

Security Information until they have 

received the appropriate authorization. 

§ 53.1118 Ineligibility of Certain 

Applicants 

This proposed rule would add 

§ 53.1118, which would prevent 

citizens, nationals, or agents of a foreign 

country or corporations owned, 

controlled, or dominated by a foreign 

entity from applying for or obtaining a 

license under part 53. 

§ 53.1120 Exceptions and Exemptions 

From Licensing Requirements 

This proposed rule would add 

§ 53.1120, which would establish the 

activities that are exempt from licensing 

requirements. 

§ 53.1121 Public Inspection of 

Applications 

This proposed rule would add 

§ 53.1121, which would allow applicant 

submissions to be made publicly 

available under the provisions of part 2. 

§ 53.1124 Relationship Between 

Sections 

This proposed rule would add 

§ 53.1124, which would outline the 

relationship between LWAs, ESPs, 

standard design approvals, standard 

DCs, MLs, CPs, OLs, and COLs under 

part 53. 

§ 53.1130 Limited Work 

Authorizations 

This proposed rule would add 

§ 53.1130, which would establish 

requirements for requesting an LWA 

and grounds for the Commission to 

issue an LWA. It would also contain 

details about the effect of an LWA and 

the implementation of a redress plan. 

§ 53.1140 Early Site Permits 

This proposed rule would add 

§ 53.1140, which would provide an 

overview of the requirements regarding 

applications for and the issuance of 

ESPs under part 53. 

§ 53.1143 Filing of Applications 

This proposed rule would add 

§ 53.1143, which would enable an 

applicant under part 53 to apply for an 

ESP, regardless of whether they have 

filed an application for a CP or COL for 

that site. 

§ 53.1144 Contents of Applications for 

Early Site Permits; General Information 

This proposed rule would add 

§ 53.1144, which would require 

applications for ESPs to include the 

information required by § 53.1109(a) 

through (d) and (j). 

§ 53.1146 Contents of Applications for 

Early Site Permits; Technical 

Information 

This proposed rule would add 

§ 53.1146, which would require 

applicants for ESPs to submit technical 

information, including but not limited 

to a Site Safety Analysis Report and 

emergency plans. 

§ 53.1149 Review of Applications 

This proposed rule would add 

§ 53.1149, which would establish 

standards for review of applications for 

ESPs under part 53, including 

requirements for the Commission to 

prepare an EIS and assess the adequacy 

of protective actions in the event of a 

radiological emergency. It would also 

require the administrative review of 

applications and hearings to follow the 

procedural requirements of part 2. 

§ 53.1155 Referral to the Advisory 

Committee on Reactor Safeguards 

This proposed rule would add 

§ 53.1115, which would require the 

ACRS to review SR content in the 

application for an ESP under part 53. 

§ 53.1158 Issuance of Early Site Permit 

This proposed rule would add 

§ 53.1158, which would establish the 

conditions under which the 

Commission may issue an ESP under 

part 53, as well as the information, 

terms, and conditions to be included in 

the permit. 

§ 53.1161 Extent of Activities 

Permitted 

This proposed rule would add 

§ 53.1161, which would require that a 

valid ESP only be used for the purpose 

of site redress, unless the site is 

referenced in an application for a CP or 

COL under part 53. 

§ 53.1164 Duration of Permit 

This proposed rule would add 

§ 53.1164, which would govern the 

conditions under which an ESP remains 

valid following the date of issuance. 

§ 53.1167 Limited Work Authorization 

After Issuance of Early Site Permit 

This proposed rule would add 

§ 53.1167, which would permit the 


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holder of an ESP to request a LWA 

under § 53.1130. 

§ 53.1170 Transfer of Early Site Permit 

This proposed rule would add 

§ 53.1170, which would govern the 

transfer of an ESP in accordance with 

§ 53.1570. 

§ 53.1173 Application for Renewal 

This proposed rule would add 

§ 53.1173, which would establish the 

conditions and procedures for renewing 

an ESP under part 53. 

§ 53.1176 Criteria for Renewal 

This proposed rule would add 

§ 53.1176, which would establish the 

criteria that the Commission may use to 

grant a renewal of an ESP under part 53. 

§ 53.1179 Duration of Renewal 

This proposed rule would add 

§ 53.1179, which would govern the 

duration of a renewed ESP under part 

53. 

§ 53.1182 Use of Site for Other 

Purposes 

This proposed rule would add 

§ 53.1182, which would govern 

acceptable uses of the site for purposes 

other than those described in the 

permit. 

§ 53.1188 Finality of Early Site Permit 

Determinations 

This proposed rule would add 

§ 53.1188, which would address the 

finality of ESP determinations under 

part 53. 

§ 53.1200 Standard Design Approvals 

This proposed rule would add 

§ 53.1200, which would address the 

procedures for filing an application for 

a standard design approval under part 

53, the process of review by NRC staff, 

and referral to the ACRS of standard 

designs. 

§ 53.1203 Filing of Applications 

This proposed rule would add 

§ 53.1203, which would enable 

applicants to submit a final design for 

the entire facility, or major portions, to 

the NRC staff for review. 

§ 53.1206 Contents of Applications for 

Standard Design Approvals; General 

Information 

This proposed rule would add 

§ 53.1206, which would require 

applications for a standard design 

approval under part 53 to contain the 

information required by § 53.1109(a) 

through (c) and (j). 

§ 53.1209 Contents of Applications for 

Standard Design Approvals; Technical 

Information 

This proposed rule would add 

§ 53.1209, which would require the 

inclusion of certain technical 

information, including a FSAR, site 

parameters, and design information, 

when an applicant seeks review of 

major portions of a standard design. 

§ 53.1210 Contents of Applications for 

Standard Design Approvals; Other 

Application Content 

This proposed rule would add 

§ 53.1210, which would require 

applications for standard design 

approvals under part 53 to include a 

description of the availability controls 

used to satisfy the safety criteria of 

§ 53.220, the program to protect 

Safeguards Information against 

unauthorized disclosure, evidence that 

safety questions associated with SSCs 

have been resolved, and a description of 

how design features fulfill design 

criteria. 

§ 53.1212 Standards for Review of 

Applications 

This proposed rule would add 

§ 53.1212, which would require 

applications for standard design 

approval to be reviewed under the 

standards in parts 20, 53, and 73. 

§ 53.1215 Referral to the Advisory 

Committee on Reactor Safeguards 

This proposed rule would add 

§ 53.1215, which would require the 

ACRS to report on any portions of the 

application for a standard design 

approval under part 53 concerning 

safety. 

§ 53.1218 Staff Approval of Design 

This proposed rule would add 

§ 53.1218, which would require the NRC 

staff to make a determination on the 

acceptability of the design, publish its 

decision in the Federal Register, and 

issue a report analyzing the design that 

is available at http://nrc.gov. 

Additionally, the rule would establish 

the conditions under which a design 

approval under part 53 remains valid. 

§ 53.1221 Finality of Standard Design 

Approvals; Information Requests 

This proposed rule would add 

§ 53.1221, which would require NRC 

staff and the ACRS to rely upon an 

approved design in their review of any 

standard DC, ML, or individual facility 

license application under part 53 that 

references the standard design approval. 

The proposed rule would also govern 

requirements for issuing information 

requests. 

§ 53.1230 Standard Design 

Certifications 

This proposed rule would add 

§ 53.1230, which would provide an 

overview of the requirements and 

procedures that govern the issuance of 

standard DCs under part 53. 

§ 53.1233 Filing of Applications 

This proposed rule would add 

§ 53.1233, which would enable an 

application for DC to be filed, regardless 

of whether an application for a CP, COL, 

or ML has been filed, provided it 

complies with the filing requirements in 

§ 53.040 and §§ 2.811 through 2.819. 

§ 53.1236 Contents of Applications for 

Standard Design Certifications; General 

Information 

This proposed rule would add 

§ 53.1236, which would require an 

application for a standard DC under part 

53 to contain all of the information 

required by § 53.1109(a) through (c) and 

(j). 

§ 53.1239 Contents of Applications for 

Standard Design Certifications; 

Technical Information 

This proposed rule would add 

§ 53.1239, which would require 

applicants for a standard DC under part 

53 to submit a FSAR that includes 

technical design information at a level 

of detail sufficient to enable the 

Commission to make a safety 

determination. 

§ 53.1241 Contents of Applications for 

Standard Design Certifications; Other 

Application Content 

This proposed rule would add 

§ 53.1241, which would require 

applications for standard DCs under 

part 53 to include an environmental 

report, as well as a description of the 

availability controls used to satisfy the 

safety criteria of § 53.220, proposed 

ITAAC, the program to protect 

Safeguards Information against 

unauthorized disclosure, evidence that 

safety questions associated with SSCs 

have been resolved, and a description of 

how design features fulfill design 

criteria. 

§ 53.1242 Review of Applications 

This proposed rule would add 

§ 53.1242, which would require 

applications for standard DCs to be 

reviewed for compliance with the 

standards in parts 20, 51, 53, and 73. It 

would also establish procedural 

requirements for reviewing applications 

and holding hearings in accordance 

with subpart H of part 2. 


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§ 53.1245 Referral to the Advisory 

Committee on Reactor Safeguards 

This proposed rule would add 

§ 53.1245, which would require the 

ACRS to report on any portions of the 

application for a standard DC under part 

53 concerning safety. 

§ 53.1248 Issuance of Standard Design 

Certification 

This proposed rule would add 

§ 53.1248, which would establish the 

conditions under which the 

Commission may issue a DC rule that 

specifies the site parameters, design 

characteristics, and any additional terms 

and conditions of the DC rule. 

§ 53.1251 Duration of Certification 

This proposed rule would add 

§ 53.1251, which would set the 

conditions under which a standard DC 

remains valid. 

§ 53.1254 Application for Renewal 

This proposed rule would add 

§ 53.1254, which would establish the 

conditions and procedures for renewing 

a standard DC under part 53. 

§ 53.1257 Criteria for Renewal 

This proposed rule would add 

§ 53.1257, which would enable the 

Commission to issue a rule granting the 

renewal of a standard DC under part 53, 

impose additional requirements, and 

grant amendment requests. 

§ 53.1260 Duration of Renewal 

This proposed rule would add 

§ 53.1260, which would provide that a 

renewal of a standard DC under part 53 

is valid for not less than 10 years, nor 

more than 15 years. 

§ 53.1263 Finality of Standard Design 

Certifications 

This proposed rule would add 

§ 53.1263, which would establish 

limited conditions under which the 

Commission may initiate a rulemaking 

to modify, rescind, or impose new 

requirements on a standard DC rule 

under part 53. It would also address 

requests for an exemption from 

elements of the certification 

information, and require that applicants 

for a CP, COL, or ML that references a 

DC rule make information normally 

contained in engineering documents 

available for audit. 

§ 53.1270 Manufacturing Licenses 

This proposed rule would add 

§ 53.1270, which would provide an 

overview of the requirements and 

procedures for applying for and issuing 

an ML under part 53. 

§ 53.1273 Filing of Applications 

This proposed rule would add 

§ 53.1273, which would establish the 

requirements to apply for an ML under 

part 53. 

§ 53.1276 Contents of Applications for 

Manufacturing Licenses; General 

Information 

This proposed rule would add 

§ 53.1276, which would require 

applicants for an ML under part 53 to 

include the information contained in 

§ 53.1109(a) through (e) and (j). 

§ 53.1279 Contents of Applications for 

Manufacturing Licenses; Technical 

Information 

This proposed rule would add 

§ 53.1279, which would require an 

applicant for an ML under part 53 to 

include certain technical information in 

a FSAR, including but not limited to 

information about site parameters, 

design information, manufacturing 

information, and information related to 

the potential fueling and ultimate 

deployment of a completed 

manufactured reactor. 

§ 53.1282 Contents of Applications for 

Manufacturing Licenses; Other 

Application Content 

This proposed rule would add 

§ 53.1282, which would require 

applicants for an ML under part 53 to 

include in their application the 

proposed ITAAC, an environmental 

report, a description of the program to 

protect Safeguards Information against 

unauthorized disclosure, and a 

description of how design features 

fulfill design criteria. It would also 

include content requirements for the 

ITAAC and environmental reports in 

applications that reference a standard 

DC. 

§ 53.1285 Review of Applications 

This proposed rule would add 

§ 53.1285, which would require 

applications for MLs under part 53 to be 

reviewed for compliance with 

applicable standards and establish 

procedural requirements for reviewing 

applicants and holding hearings in 

accordance with part 2. 

§ 53.1286 Referral to the Advisory 

Committee on Reactor Safeguards 

This proposed rule would add 

§ 53.1286, which would require the 

ACRS to report on any portions of the 

application for an ML under part 53 

concerning safety. 

§ 53.1287 Issuance of Manufacturing 

Licenses 

This proposed rule would add 

§ 53.1287, which would establish the 

conditions under which the 

Commission may issue an ML under 

part 53. 

§ 53.1288 Finality of Manufacturing 

Licenses 

This proposed rule would add 

§ 53.1288, which would address the 

limited circumstances in which the 

Commission may modify, rescind, or 

impose new requirements following the 

issuance of an ML under part 53. It 

would also address requests for a 

departure from the specifications of the 

license. 

§ 53.1291 Duration of Manufacturing 

Licenses 

This proposed rule would add 

§ 53.1291, which would govern the 

expiration of an ML, which is valid for 

no less than 5, nor more than 15 years 

from the date of issuance. 

§ 53.1293 Transfer of Manufacturing 

Licenses 

This proposed rule would add 

§ 53.1293, which would provide that an 

ML under part 53 may be transferred in 

accordance with § 53.1570. 

§ 53.1295 Renewal of Manufacturing 

Licenses 

This proposed rule would add 

§ 53.1295, which would establish the 

procedures for applicants to apply for 

and the Commission to grant a renewal 

of an ML under part 53. 

§ 53.1300 Construction Permits 

This proposed rule would add 

§ 53.1300, which would provide an 

overview of the requirements and 

procedures for applicants to apply for 

and the Commission to grant a CP under 

part 53. 

§ 53.1306 Contents of Applications for 

Construction Permits; General 

Information 

This proposed rule would add 

§ 53.1306, which would require 

applicants for a CP under part 53 to 

submit the general information required 

by § 53.1109, as well as financial 

information. 

§ 53.1309 Contents of Applications for 

Construction Permits; Technical 

Information 

This proposed rule would add 

§ 53.1309, which would require 

applicants for a CP under part 53 to 

submit a PSAR and a description of the 

program to protect Safeguards 


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Information from unauthorized 

disclosure. 

§ 53.1312 Contents of Applications for 

Construction Permits; Other Application 

Content 

This proposed rule would add 

§ 53.1312, which would require 

applicants for a CP under part 53 to 

submit an environmental report and to 

provide additional details in the PSAR 

if the application references an ESP, 

standard design approval, or standard 

DC. 

§ 53.1315 Review of Applications 

This proposed rule would add 

§ 53.1315, which would require 

applications for CPs under part 53 to be 

reviewed for compliance with 

applicable standards and establish 

procedural requirements for reviewing 

applications and holding hearings in 

accordance with part 2. 

§ 53.1318 Finality of Referenced NRC 

Approvals, Permits, and Certifications 

This proposed rule would add 

§ 53.1318, which would address the 

finality of ESPs, standard design 

approvals, and standard DCs referenced 

in the CP application. 

§ 53.1324 Referral to the Advisory 

Committee on Reactor Safeguards 

This proposed rule would add 

§ 53.1324, which would require the 

ACRS to report on any portions of the 

application for a CP under part 53 

concerning safety. 

§ 53.1327 Authorization To Conduct 

Limited Work Authorization Activities 

This proposed rule would add 

§ 53.1327, which would govern 

authorization to conduct LWA 

activities. 

§ 53.1330 Exemptions, Departures, 

and Variances 

This proposed rule would add 

§ 53.1330, which would govern requests 

for and issuance of exemptions from the 

Commission’s regulations and 

exemptions, departures, and variances 

from NRC approvals, permits, and 

certifications. 

§ 53.1333 Issuance of Construction 

Permits 

This proposed rule would add 

§ 53.1333, which would establish the 

conditions under which the 

Commission may issue CPs and 

accompanying terms and conditions 

under part 53. 

§ 53.1336 Finality of Construction 

Permits 

This proposed rule would add 

§ 53.1336, which would address the 

finality of CPs. 

§ 53.1342 Duration of Construction 

Permits 

This proposed rule would add 

§ 53.1342, which would establish 

requirements for the expiration of a CP. 

§ 53.1345 Transfer of Construction 

Permits 

This proposed rule would add 

§ 53.1345, which would govern the 

transfer of CPs under part 53. 

§ 53.1348 Termination of Construction 

Permits 

This proposed rule would add 

§ 53.1348, which would require the 

holder of a permit under part 53 to 

provide written certification to the 

Commission within 30 days of 

determining to permanently cease 

construction. 

§ 53.1360 Operating Licenses 

This proposed rule would add 

§ 53.1360, which would provide an 

overview of the requirements and 

procedures for applicants to apply for 

and the Commission to issue an OL 

under part 53. 

§ 53.1366 Contents of Applications for 

Operating Licenses; General Information 

This proposed rule would add 

§ 53.1366, which would require an 

application for an OL under part 53 to 

include the information required by 

§ 53.1109 as well as financial 

information. 

§ 53.1369 Contents of Applications for 

Operating Licenses; Technical 

Information 

This proposed rule would add 

§ 53.1369, which would require an 

application for an OL under part 53 to 

include certain technical information in 

an FSAR at a level of detail sufficient for 

the Commission to reach a final 

conclusion on all safety matters. 

§ 53.1372 Contents of Applications for 

Operating Licenses; Other Application 

Content 

This proposed rule would add 

§ 53.1372, which would require an 

application for an OL under part 53 to 

include an environmental report and a 

description of availability controls. 

§ 53.1375 Review of Applications 

This proposed rule would add 

§ 53.1375, which would establish the 

standards and procedures for reviewing 

applications and holding hearings on 

OLs under part 53. 

§ 53.1381 Referral to the Advisory 

Committee on Reactor Safeguards 

This proposed rule would add 

§ 53.1381, which would require the 

ACRS to report on any portions of the 

application for a CP under part 53 

concerning safety. 

§ 53.1384 Exemptions, Departures, 

and Variances 

This proposed rule would add 

§ 53.1384, which would govern requests 

for and the issuance of exemptions from 

the Commission’s regulations and 

exemptions, departures, and variances 

from NRC approvals, permits, and 

certifications. 

§ 53.1387 Issuance of Operating 

Licenses 

This proposed rule would add 

§ 53.1387, which would establish the 

conditions under which the 

Commission may issue OLs and 

accompanying conditions and 

limitations, including TS, under part 53. 

§ 53.1390 Backfitting of Operating 

Licenses 

This proposed rule would add 

§ 53.1390, which would prevent the 

Commission from modifying, adding, or 

deleting any terms or conditions of the 

OL, except in accordance with 

§ 53.1590. 

§ 53.1396 Duration of Operating 

Licenses 

This proposed rule would add 

§ 53.1396, which would provide that an 

OL under part 53 may be valid for up 

to 40 years. 

§ 53.1399 Transfer of an Operating 

License 

This proposed rule would add 

§ 53.1399, which would provide that an 

OL under part 53 may be transferred 

under § 53.1570. 

§ 53.1402 Application for Renewal 

This proposed rule would add 

§ 53.1402, which would provide that an 

application for a renewed OL under part 

53 must be filed in accordance with 

§ 53.1595. 

§ 53.1405 Continuation of an 

Operating License 

This proposed rule would add 

§ 53.1405, which would govern the 

continuing obligations of the holder of 

an OL under part 53 following the 

permanent cessation of operations. 


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§ 53.1410 Combined Licenses 

This proposed rule would add 

§ 53.1410, which would provide an 

overview of the requirements and 

procedures for applicants to apply for 

and the Commission to issue a COL 

under part 53. 

§ 53.1413 Contents of Applications for 

Combined Licenses; General 

Information 

This proposed rule would add 

§ 53.1413, which would require an 

application for a COL under part 53 to 

include the information required by 

§ 53.1109 as well as financial 

information. 

§ 53.1416 Contents of Applications for 

Combined Licenses; Technical 

Information 

This proposed rule would add 

§ 53.1416, which would require 

applicants for a COL under part 53 to 

submit an FSAR with a level of 

technical information sufficient to reach 

a final conclusion on all safety matters. 

§ 53.1419 Contents of Applications for 

Combined Licenses; Other Application 

Content 

This proposed rule would add 

§ 53.1419, which would require 

applicants for a COL under part 53 to 

submit an environmental report, a 

description of availability controls, the 

ITAAC that the licensee must perform. 

It would also include ITAAC 

requirements for applications that 

reference an ESP, standard DC, ML, or 

combination thereof. 

§ 53.1422 Review of Applications 

This proposed rule would add 

§ 53.1422, which would require 

applications for COLs under part 53 to 

be reviewed for compliance with 

applicable standards and establish 

procedural requirements for reviewing 

applications and holding hearings in 

accordance with part 2. 

§ 53.1425 Finality of Referenced NRC 

Approvals 

This proposed rule would add 

§ 53.1425 which would address the 

finality of ESPs, standard DC rules, 

standard design approvals, or MLs 

referenced in the application for a COL 

under part 53. 

§ 53.1431 Referral to the Advisory 

Committee on Reactor Safeguards 

This proposed rule would add 

§ 53.1431, which would require the 

ACRS to report on any portions of the 

application for a COL under part 53 

concerning safety. 

§ 53.1434 Authorization To Conduct 

Limited Work Authorization Activities 

This proposed rule would add 

§ 53.1434, which would address 

authorization to conduct LWA 

activities. 

§ 53.1437 Exemptions, Departures, 

and Variances 

This proposed rule would add 

§ 53.1437, which would govern the 

conditions in which the Commission 

may grant an exemption for one or more 

of its regulations, or an exemption, 

variance, or departure from a permit, 

design approval, or license. 

§ 53.1440 Issuance of Combined 

Licenses 

This proposed rule would add 

§ 53.1440, which would establish the 

conditions under which the 

Commission may issue COLs and 

accompanying conditions and 

limitations, including TS, under part 53. 

§ 53.1443 Finality of Combined 

Licenses 

This proposed rule would add 

§ 53.1443, which would govern 

permissible modifications or 

amendments that the Commission may 

make to a COL, as well as permissible 

changes that a licensee may make to 

facilities and procedures as described in 

the FSAR. 

§ 53.1449 Inspection During 

Construction 

This proposed rule would add 

§ 53.1449, which would establish 

requirements related to inspections, 

tests, or analyses for the holder of a COL 

under part 53. 

§ 53.1452 Operation Under a 

Combined License 

This proposed rule would add 

§ 53.1452, which would establish 

requirements describing the 

notifications, hearings, and findings to 

be made prior to commencing facility 

operations. 

§ 53.1455 Duration of a Combined 

License 

This proposed rule would add 

§ 53.1455, which would govern the 

duration of a COL under part 53. 

§ 53.1456 Transfer of a Combined 

License 

This proposed rule would add 

§ 53.1456, which would permit the 

transfer of a COL under part 53 in 

accordance with § 53.1570. 

§ 53.1458 Application for Renewal 

This proposed rule would add 

§ 53.1458, which would provide that an 

application for renewal of a COL must 

be filed in accordance with § 53.1595. 

§ 53.1461 Continuation of Combined 

License 

This proposed rule would add 

§ 53.1461, which would govern the 

continuing obligations of the holder of 

a COL under part 53 following the 

permanent cessation of operations. 

§ 53.1470 Standardization of 

Commercial Nuclear Plant Designs: 

Licenses To Construct and Operate 

Nuclear Power Reactors of Identical 

Design at Multiple Sites 

This proposed rule would add 

§ 53.1470, which would govern the 

requirements and procedures for filing 

and issuing applications for a CP, OL, or 

COL under part 53 in which the 

applicant seeks approval of the same 

design for multiple sites. 

Subpart I—Maintaining and Revising 

Licensing-Basis Information 

This proposed rule would add subpart 

I, which would address the maintenance 

of licensing-basis information for part 

53. 

§ 53.1500 Licensing-Basis Information 

This proposed rule would add 

§ 53.1500, describing the purpose of 

subpart I, which would be to provide 

the requirements for the maintenance of 

licensing-basis information for 

commercial nuclear plants licensed 

under part 53. 

§ 53.1502 Specific Terms and 

Conditions of Licenses 

This proposed rule would add 

§ 53.1502, which would outline the 

specific terms and conditions for 

obtaining a license under part 53. 

§ 53.1505 Changes to Licensing-Basis 

Information Requiring Prior NRC 

Approval 

This proposed rule would add 

§ 53.1505, which would provide an 

overview of the process for licensees to 

request, and the Commission to issue, 

amendments to licensing-basis 

information under part 53. 

§ 53.1510 Application for Amendment 

of License 

This proposed rule would add 

§ 53.1510, which would require 

licensees under part 53 to file an 

application to request an amendment to 

the license. Applicants must assess how 

their requested changes would impact 

the safety criteria and analysis 


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requirements in subpart B and C, as 

applicable, whether the amendment 

involves no significant hazards 

consideration using the standards in 

§ 53.1520 and consider potential 

impacts on environmental factors. 

§ 53.1515 Public Notices; State 

Consultation 

This proposed rule would add 

§ 53.1515, which would outline the 

Commission’s procedures for issuing a 

notification in the Federal Register and 

consulting with the State in which the 

commercial nuclear facility is located in 

connection with its consideration of 

applications for an amendment to an OL 

or COL under part 53. 

§ 53.1520 Issuance of Amendment 

This proposed rule would add 

§ 53.1520, which would outline criteria 

for the Commission to consider in 

issuing license amendments under part 

53. 

§ 53.1525 Revising Certification 

Information Within a Design 

Certification Rule 

This proposed rule would add 

§ 53.1525, which would address the 

requirements for applicants to request, 

and the Commission to grant, an 

exemption to a DC rule under part 53. 

§ 53.1530 Revising Design Information 

Within a Manufacturing License 

This proposed rule would add 

§ 53.1530, which would require the 

holder of an ML to request an 

amendment under § 53.1510 and, as 

applicable, § 53.1520 to make changes to 

the design of a manufactured reactor. It 

would also outline the requirements for 

holders of a COL under part 53 to 

request amendments for changes to the 

design information of a manufactured 

reactor. 

§ 53.1535 Amendments During 

Construction 

This proposed rule would add 

§ 53.1535, which would outline the 

process for licensees under part 53 to 

request amendments to CPs or LWAs 

during construction. 

§ 53.1540 Updating Licensing-Basis 

Information and Determining the Need 

for NRC Approval 

This proposed rule would add 

§ 53.1540, which would provide an 

overview of the regulations in subpart I 

for holders of an OL or COL under part 

53 to modify licensing-basis information 

and definitions relevant to §§ 53.1545 

through 53.1565. 

§ 53.1545 Updating Final Safety 

Analysis Reports 

This proposed rule would add 

§ 53.1545, which would require 

licensees under part 53 to regularly 

update FSARs in accordance with the 

requirements of this section to reflect 

changes to licensing-basis information. 

§ 53.1550 Evaluating Changes to 

Facility as Described in Final Safety 

Analysis Reports 

This proposed rule would add 

§ 53.1550, which would require 

licensees under part 53 to follow the 

guidelines outlined in this section in 

determining whether changes to 

licensing-basis information described in 

the FSAR (as updated) require them to 

obtain a license amendment. 

§ 53.1560 Updating Program 

Documents Included in Licensing-Basis 

Information 

This proposed rule would add 

§ 53.1560, which would require the 

holders of an OL or COL under part 53 

to regularly update the program 

documents that they submitted in their 

application for a license. 

§ 53.1565 Evaluating Changes to 

Programs Included in Licensing-Basis 

Information 

This proposed rule would add 

§ 53.1565, which would enable 

licensees under part 53 to make changes 

to the facility, procedures, or 

organization, or address changes to site 

environs as described in program 

documents without NRC approval if 

these changes satisfy the criteria 

outlined in this section. 

§ 53.1570 Transfer of Licenses 

This proposed rule would add 

§ 53.1570, which would outline the 

requirements for an application for 

transfer of a license issued under part 

53. 

§ 53.1575 Termination of Licenses 

This proposed rule would add 

§ 53.1575, which would outline the 

process for terminating an OL or COL 

issued under part 53. 

§ 53.1580 Information Requests 

This proposed rule would add 

§ 53.1580, which would address the 

process and circumstances under which 

the NRC may send information requests 

to the various types of licensees within 

part 53. 

§ 53.1585 Revocation, Suspension, 

Modification of Licenses and Approvals 

for Cause 

This proposed rule would add 

§ 53.1585, which would address 

grounds for the revocation, suspension, 

or modification of a license or standard 

design approval issued under part 53. 

§ 53.1590 Backfitting 

This proposed rule would add 

§ 53.1590, which would define 

backfitting and establish requirements 

to be met by the NRC when it takes 

backfitting actions under part 53. 

§ 53.1595 Renewal 

This proposed rule would add 

§ 53.1595, which would provide for the 

renewal of a license under part 53 upon 

expiration. 

Subpart J—Reporting and Other 

Administrative Requirements 

This proposed rule would add subpart 

J, to establish various reporting and 

other administrative requirements for 

licensees under part 53. 

§ 53.1600 General Information 

This proposed rule would add 

§ 53.1600, which provides an overview 

of the sections that would require 

applicants and licensees under part 53 

to provide NRC inspectors with 

unfettered access to sites and facilities, 

maintain records and make reports, 

demonstrate compliance with financial 

qualification and reporting 

requirements, and maintain required 

financial protection for accidents. 

§ 53.1610 Unfettered Access for 

Inspections 

This proposed rule would add 

§ 53.1610, which would require 

applicants and licensees under part 53 

to provide unfettered access to NRC 

inspectors, including access to records, 

premises, activities, and licensed 

materials, in addition to providing office 

space to accommodate temporary or 

resident inspectors. 

§ 53.1620 Maintenance of Records, 

Making of Reports 

This proposed rule would add 

§ 53.1620, which would require part 53 

licensees to maintain all records and 

make reports as required by the 

conditions of the license or by the 

regulations in part 53. 

§ 53.1630 Immediate Notification 

Requirements for Operating Commercial 

Nuclear Plants 

This proposed rule would add 

§ 53.1630, which would impose 

immediate notification requirements on 


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part 53 licensees following the 

declaration of an Emergency Class or the 

discovery of certain non-emergency 

events. 

§ 53.1640 Licensee Event Report 

System 

This proposed rule would add 

§ 53.1640, which would require any 

commercial plant licensee holding an 

OL under part 53 to submit a Licensee 

Event Report in accordance with the 

specifications outlined in this section. 

§ 53.1645 Reports of Radiation 

Exposure to Members of the Public 

The proposed rule would add 

§ 53.1645, which would require annual 

reports to the Commission, including 

radiological reports as required by part 

20, an Annual Radioactive Effluent 

Release Report, and an Annual 

Environmental Operating Report. 

§ 53.1650 Facility Information and 

Verification 

The proposed rule would add 

§ 53.1650, which would include a 

reporting requirement for applicants 

and holders of a CP or license under 

part 53 to support safeguards 

agreements between the United States 

and the IAEA. 

§ 53.1660 Financial Requirements 

This proposed rule would add 

§ 53.1660, which would introduce 

requirements and procedures related to 

financial qualifications and reporting 

requirements for applicants, licensees, 

and CP holders under part 53. 

§ 53.1670 Financial Qualifications 

This proposed rule would add 

§ 53.1670, which would require an 

applicant for a CP, OL, or COL under 

part 53 to must demonstrate possession 

or ability to obtain funds necessary for 

the activities for which the permit or 

license is sought. 

§ 53.1680 Annual Financial Reports 

This proposed rule would add 

§ 53.1680, which would require 

licensees and holders of a CP under part 

53 to submit annual financial reports to 

the Commission, with exceptions for 

those that submit financial forms to the 

Securities and Exchange Commission or 

the Federal Energy Regulatory 

Commission. 

§ 53.1690 Licensee’s Change of Status; 

Financial Qualifications 

This proposed rule would add 

§ 53.1690, which would require electric 

utility licensees that hold an OL or COL 

for a commercial nuclear plant under 

part 53 to provide the NRC with the 

financial qualifications information 

outlined in this section within seventyfive days of ceasing to be an electric 

utility. 

§ 53.1700 Creditor Regulations 

This proposed rule would add 

§ 53.1700, which would establish 

regulations with respect to the creditors 

of any facility under part 53. 

§ 53.1710 Financial Protection 

This proposed rule would add 

§ 53.1710, which would establish 

requirements for licenses under part 53 

to obtain and maintain insurance to 

cover the costs of an accident. 

§ 53.1720 Insurance Required To 

Stabilize and Decontaminate Plant 

Following an Accident 

This proposed rule would add 

§ 53.1720, which would require 

commercial nuclear plant licensees 

under part 53 to obtain insurance 

sufficient to cover the costs of 

stabilizing and decontaminating the 

plant in the event of an accident. 

§ 53.1730 Financial Protection 

Requirements 

This proposed rule would add 

§ 53.1730, which would require 

commercial nuclear plant licensees 

under part 53 to satisfy the provisions 

of part 140. 

Subpart M—Enforcement 

This proposed rule would add subpart 

M, which would address certain 

violations and penalties associated with 

violations of part 53 regulations. 

§ 53.9000 Violations 

This proposed rule would add 

§ 53.9000, providing notice of the 

Commission’s authority to obtain 

injunctions or other court orders for the 

violations enumerated in this section. 

§ 53.9010 Criminal Penalties 

This proposed rule would add 

§ 53.9010, providing notice to all 

persons and entities subject to part 53 

that they are subject to criminal 

sanctions for willful violations, 

attempted violations, or conspiracy to 

violate certain regulations under part 

53. 

§ 70.20a General License to Possess 

Special Nuclear Material for Transport 

This proposed rule would revise 

§ 70.20a(b) to include a reference to part 

53. 

§ 70.22 Contents of Applications 

This proposed rule would revise 

§ 70.22, paragraphs (b), (h)(1), (j)(1), and 

(k) to include the appropriate references 

to part 53. 

§ 70.24 Criticality Accident 

Requirements 

This proposed rule would revise 

§ 70.24(d) to include the appropriate 

references to part 53. 

§ 70.32 Conditions of Licenses 

This proposed rule would revise 

§ 70.32(c)(1) and (d) to incorporate the 

appropriate references to part 53. 

§ 70.50 Reporting Requirements 

This proposed rule would revise 

§ 70.50(d) to clarify the applicability of 

the reporting requirements of this 

section to part 53 licensees. 

§ 72.3 Definitions 

This proposed rule would revise the 

definition of ‘‘Independent spent fuel 

storage installation or ISFSI’’ in § 72.3 to 

include a reference to facilities licensed 

under part 53. 

§ 72.30 Financial Assurance and 

Recordkeeping for Decommissioning 

This proposed rule would revise 

§ 72.30(e)(5) to include the appropriate 

references to part 53. 

§ 72.32 Emergency Plan 

This proposed rule would revise 

§ 72.32(c)(2) to include a reference to 

the exclusion area as defined in part 53. 

§ 72.40 Issuance of License 

This proposed rule would revise 

§ 72.40(c) regarding the issuance of a 

license under part 72 to include a 

reference to previous licensing actions, 

including the issuance of a CP under 

part 53. 

§ 72.75 Reporting Requirements for 

Specific Events and Conditions 

This proposed rule would revise 

§ 72.75(i)(1)(ii) regarding reporting 

requirements for specific events and 

conditions with references to reactors 

licensed under part 53. 

§ 72.184 Safeguards Contingency Plan 

This proposed rule would revise 

§ 72.184(a) regarding the requirements 

of a licensee’s safeguarding contingency 

plan with a reference to nuclear 

facilities licensed under part 53. 

§ 72.210 General License Issued 

This proposed rule would revise 

§ 72.210 to issue a general license for 

the storage of spent fuel in an 

independent spent storage installation 

at power to persons authorized to 

possess or operate nuclear power 

reactors under part 53. 


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§ 72.212 Conditions of General License 

Issued Under § 72.210 

This proposed rule would revise 

§ 72.212(b)(8) regarding the conditions 

of a general license issued under 

§ 72.210 to include a reference to license 

amendments for a facility made 

pursuant to part 53. 

§ 72.218 Termination of Licenses 

This proposed rule would revise 

§ 72.218(a) to include a reference to the 

notification required under part 53 

regarding the plan for managing spent 

fuel prior to decommissioning. It would 

also extend the provisions of § 72.218(b) 

to a reactor operating or COL under part 

53. 

§ 73.1 Purpose and Scope 

This proposed rule would revise 

§ 73.1(b)(1)(i) to extend the scope of part 

73 to production and utilization 

facilities licensed under part 53, in 

addition to parts 50 and 52. 

§ 73.2 Definitions 

This proposed rule would revise 

§ 73.2 introductory text and paragraph 

(a) such that terms defined in part 53 

have the same meaning in part 73. 

§ 73.8 Information Collection 

Requirements: OMB Approval 

This proposed rule would revise 

§ 73.8(b) with the new information 

collection requirements contained in 

proposed §§ 73.77, 73.100, 73.110, and 

73.120. 

§ 73.50 Requirements for Physical 

Protection of Licensed Activities 

This proposed rule would revise 

§ 73.50 to exempt nuclear reactor 

facilities licensed under part 53, in 

addition to parts 50 and 52, from the 

requirements of this section. 

§ 73.55 Requirements for Physical 

Protection of Licensed Activities in 

Nuclear Power Reactors Against 

Radiological Sabotage 

This proposed rule would revise 

§ 73.55, paragraphs (a)(4) and (6), 

(i)(4)(iii), (l)(1), (l)(7)(ii), (p)(1)(i), (r)(2), 

and (r)(4)(iii), to incorporate the 

appropriate references to part 53 

regarding requirements for physical 

protection of licensed activities in 

nuclear power reactors against 

radiological sabotage. 

§ 73.56 Personnel Access 

Authorization Requirements for Nuclear 

Power Plants 

This proposed rule would revise 

§ 73.56(a)(3) to apply this section’s 

personnel AA requirements to 

applicants for an OL or holders of a COL 

under part 53 who do not demonstrate 

compliance with certain requirements 

under part 53. 

§ 73.57 Requirements for Criminal 

History Records Checks of Individuals 

Granted Unescorted Access to a Nuclear 

Power Facility, a Non-power Reactor, or 

Access to Safeguards Information 

This proposed rule would revise 

§ 73.57(a)(3) to incorporate the 

appropriate references to OLs granted 

under part 53 and Commission findings 

under § 53.1452(g) regarding the 

requirement for license applicants to 

submit fingerprints for all personnel 

with unescorted access. 

§ 73.58 Safety/Security Interface 

Requirements for Nuclear Power 

Reactors 

This proposed rule would revise 

§ 73.58(a) to extend the requirements of 

this section to part 53 licensees. 

§ 73.67 Licensee Fixed Site and InTransit Requirements for the Physical 

Protection of Special Nuclear Material 

of Moderate and Low Strategic 

Significance 

This proposed rule would revise 

§ 73.67(d) and (f) to include a reference 

to licensees authorized to operate a 

nuclear power plant under part 53. 

§ 73.77 Cybersecurity Event 

Notifications 

This proposed rule would revise 

§ 73.77, paragraphs (a), (b), (c)(6) and (7) 

regarding the notification process for 

cybersecurity events to include 

notifications for the declaration of an 

emergency class made under part 53. 

Subpart J—Security Requirements at 

Commercial Nuclear Plants 

This proposed rule would add new 

Subpart J of part 73 containing 

§§ 73.100, 73.110, and 73.120, to 

establish security requirements for 

commercial nuclear plants licensed 

under part 53. 

§ 73.100 Technology-Inclusive 

Requirements for Physical Protection of 

Licensed Activities at Commercial 

Nuclear Plants Against Radiological 

Sabotage 

This proposed rule would add 

§ 73.100, which would establish a 

performance-based regulatory 

framework for physical protection as an 

alternative to the prescriptive 

requirements of § 73.55, which also 

governs physical protection programs 

for part 50 and 52 licensees. 

§ 73.110 Technology-Inclusive 

Requirements for Protection of Digital 

Computer and Communication Systems 

and Networks 

This proposed rule would add 

§ 73.110, which would establish a 

consequence-based approach to 

cybersecurity and would require that 

part 53 licensees demonstrate 

reasonable assurance that digital 

computer and communication systems 

and networks are adequately protected 

against cyberattacks in a manner that is 

commensurate with the potential 

consequences of those attacks. 

§ 73.120 Access Authorization 

Program for Commercial Nuclear Plants 

This proposed rule would add 

§ 73.120, which would establish 

performance objectives as an alternative 

to compliance with the AA provisions 

of §§ 73.55, 73.56, and 73.57. This 

proposed rule would afford part 53 

licensees additional flexibility in 

establishing an AA program that 

demonstrates compliance with the 

performance objectives and 

requirements of this section. 

§ 73.1200 Notification of Physical 

Security Events 

This proposed rule would revise 

§ 73.1200, paragraphs (a), (c)(1), (e)(1), 

(e)(3), (e)(4), (g)(1), (o)(5)(i), (o)(6)(i), (r), 

and (s) to extend the requirements of 

this section to part 53 licensees. 

§ 73.1205 Written Follow-Up Reports 

of Physical Security Events 

This proposed rule would revise 

§ 73.1205(b)(2) to extend the 

requirements of this section to part 53 

licensees. 

§ 73.1210 Recordkeeping of Physical 

Security Events 

This proposed rule would revise 

§ 73.1210(a)(1) and (b)(3)(i) to extend 

the requirements of this section to part 

53 licensees. 

§ 73.1215 Suspicious Activity Reports 

This proposed rule would revise 

§ 73.1215(d)(1) to include a reference to 

§ 73.100. 

Appendix B to part 73—General Criteria 

for Security Personnel 

This proposed rule would revise 

appendix B to part 73 to state that terms 

defined in part 53 have the same 

meaning when used in this appendix. 

§ 74.31 Nuclear Material Control and 

Accounting for Special Nuclear Material 

of Low Strategic Significance 

This proposed rule would revise 

§ 74.31(a) to include a reference to 


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production or utilization facilities 

licensed under part 53, in addition to 

parts 50 and 70. 

§ 74.41 Nuclear Material Control and 

Accounting for Special Nuclear Material 

of Moderate Strategic Significance 

This proposed rule would revise 

§ 74.41(a) to include a reference to 

nuclear reactors licensed under part 53. 

§ 74.51 Nuclear Material Control and 

Accounting for Strategic Special 

Nuclear Material 

This proposed rule would revise 

§ 74.51(a) to include a reference to 

nuclear reactors licensed under part 53. 

§ 75.4 Definitions 

This proposed rule would revise 

§ 75.4 such that terms defined in 

§ 53.020 have the same meaning when 

used in this part. The definition of 

‘‘Facility’’ would also be revised to 

include any plant or location where 

more than 1 effective kilogram of 

nuclear material is licensed pursuant to 

part 53. 

§ 95.5 Definitions 

This proposed rule would revise the 

definition of ‘‘License’’ in § 95.5 to 

include those issued under part 53. 

§ 95.39 External Transmission of 

Documents and Material 

This proposed rule would revise 

§ 95.39(a) to apply restrictions to the 

external transmission of documents and 

material containing classified 

information in connection with NRC 

licenses, certificates, standard design 

approvals, or standard DCs issued under 

part 53. 

§ 140.2 Scope 

This proposed rule would revise 

§ 140.2(a)(1) and (2) to include part 53 

applicants and licensees within the 

scope of part 140 regulations. 

§ 140.10 Scope 

This proposed rule would revise 

§ 140.10 to apply the provisions of 

subpart B to applicants or holders of a 

license to operate a nuclear reactor 

under part 53, as well as applicants and 

holders of a COL under part 53. 

§ 140.11 Amounts of Financial 

Protection for Certain Reactors 

This proposed rule would revise 

§ 140.11(b) to require the licensee’s 

primary financial protection to cover all 

reactors in any case where a person is 

authorized under part 53 to operate two 

or more nuclear reactors at the same 

location. 

§ 140.12 Amount of Financial 

Protection Required for Other Reactors 

This proposed rule would revise 

§ 140.12(c) to require the licensee’s 

primary financial protection to cover all 

reactors in any case where a person is 

authorized under part 53 to operate two 

or more nuclear reactors at the same 

location. 

§ 140.13 Amount of Financial 

Protection Required of Certain Holders 

of Construction Permits and Combined 

Licenses Under 10 CFR Part 52 

This proposed rule would revise 

§ 140.13 with the appropriate references 

to part 53 regarding the requirement for 

holders of a CP or COL under part 53 

to obtain financial protection. 

§ 140.20 Indemnity Agreements and 

Liens 

This proposed rule would revise 

§ 140.20(a)(1)(i) and (ii) with 

appropriate references to part 53. 

§ 150.15 Persons Not Exempt 

The proposed rule would revise 

§ 150.15, paragraphs (a)(7)(iii) and (a)(8) 

to add a reference to facilities licensed 

under parts 53 and 52. 

§ 170.3 Definitions 

The proposed rule would revise 

§ 170.3 to incorporate references to part 

53 into the definitions of 

‘‘Manufacturing license,’’ ‘‘Part 55 

Reviews,’’ ‘‘Power reactor,’’ and 

‘‘Special projects.’’ 

§ 170.12 Payment of Fees 

The proposed rule would revise 

§ 170.12(d)(1)(v) regarding special 

project fees in connection with FSARs 

to include part 53. 

§ 170.21 Schedule of Fees for 

Production and Utilization Facilities, 

Review of Standard Referenced Design 

Approvals, Special Projects, 

Inspections, And import and Export 

Licenses 

The proposed rule would revise 

§ 170.21, footnote 1 to include fees 

charged for approvals issued under the 

exemption provision in § 53.080. 

§ 170.41 Failure by Applicant or 

Licensee to Pay Prescribed Fees 

The proposed rule would revise 

§ 170.41 to include a general reference 

to part 53 in connection with remedial 

actions that the Commission might take 

when an applicant or licensee fails to 

pay a prescribed fee required by this 

part. 

§ 171.3 Scope 

The proposed rule would revise 

§ 171.3 to apply the provisions of this 

part to any person holding an OL for a 

power reactor licensed under part 53 or 

a COL issued under part 53. 

§ 171.5 Definitions 

This proposed rule would revise the 

definitions of ‘‘Operating license’’ and 

‘‘Power reactor’’ in § 171.5 to 

incorporate the appropriate references 

to part 53. 

§ 171.15 Annual fees: Non-Power 

Production or Utilization Licenses, 

Reactor Licenses, and Independent 

Spent Fuel Storage Licenses 

This proposed rule would revise 

§ 171.15, paragraphs (a), (b)(2)(iii), 

(c)(1), and (d)(1) regarding annual fees 

that are applicable to part 53 licensees. 

§ 171.17 Proration 

This proposed rule would revise 

§ 171.17, paragraphs (a), (a)(1)(ii) and 

(a)(2) with references to part 53 licenses. 

VIII. Regulatory Flexibility 

Certification 

The Regulatory Flexibility Act of 

1980, as amended at 5 U.S.C. 601 et seq, 

requires that agencies consider the 

impact of their rulemakings on small 

entities and, consistent with applicable 

statutes, consider alternatives to 

minimize these impacts on the 

businesses, organizations, and 

government jurisdictions to which they 

apply. 

In accordance with the Small 

Business Administration’s (SBA’s) 

regulation at 13 CFR 121.903(c), the 

NRC has developed its own size 

standards for performing an RFA 

analysis and has verified with the SBA 

Office of Advocacy that its size 

standards are appropriate for NRC 

analyses. The NRC size standards at 

§ 2.810, ‘‘NRC size standards,’’ are used 

to determine whether an applicant or 

licensee qualifies as a small entity in the 

NRC’s regulatory programs. Section 

2.810 defines the following types of 

small entities: 

Small business is a for-profit concern 

and is a—(1) Concern that provides a 

service or a concern not engaged in 

manufacturing with average gross 

receipts of $8.0 million or less over its 

last 5 completed fiscal years; or (2) 

Manufacturing concern with an average 

number of 500 or fewer employees 

based upon employment during each 

pay period for the preceding 12 calendar 

months. 

Small organization is a not-for-profit 

organization which is independently 


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owned and operated and has annual 

gross receipts of $8.0 million or less. 

Small governmental jurisdiction is a 

government of a city, county, town, 

township, village, school district, or 

special district with a population of less 

than 50,000. 

Small educational institution is one 

that is—(1) Supported by a qualifying 

small governmental jurisdiction; or (2) 

Not State or publicly supported and has 

500 or fewer employees. 

Number of Small Entities Affected 

The NRC is currently not aware of any 

known small entities as defined in 

§ 2.810 that are planning to apply for a 

commercial nuclear plant ESP, CP, OL, 

ML, or COL under part 53 that would be 

impacted by this proposed rule. Based 

on this finding, the NRC has 

preliminarily determined that the 

proposed rule would not have a 

significant economic impact on a 

substantial number of small entities. 

Economic Impact on Small Entities 

Depending on how the ownership 

and/or operating responsibilities for 

such an enterprise were structured, 

applicants for a commercial nuclear 

plant rated 8 Megawatts electric (MWe) 

or less could conceivably qualify as 

small entities as defined by § 2.810. 

Owners that operate power reactors 

rated greater than 8 MWe could generate 

sufficient electricity revenue that 

exceeds the gross annual receipts limit 

of $8 million, assuming a 90 percent 

capacity factor and the June 2021 DOE’s 

Energy Information Administration U.S. 

average price of electricity to the 

ultimate customer for all sectors of 11.3 

cents per kilowatt-hour. 

Although the NRC is not aware of any 

small entities that would be affected by 

the proposed rule, there is a possibility 

that future applications for a 

commercial nuclear plant permit or 

license could be submitted by small 

entities who plan to own and operate a 

commercial nuclear plant rated 8 MWe 

or less. Commercial nuclear plants that 

are rated 8 MWe or less would most 

likely be used to support electrical 

demand for military bases or small 

remote towns and would provide 

process heat, so they would not directly 

compete with a larger commercial 

nuclear plant that would typically 

produce electricity for the grid. As a 

result of these differing purposes, the 

NRC would expect that small and large 

entities would not be in direct 

competition with each other. 

Therefore, the NRC preliminarily 

concludes that this proposed rule would 

not have a significant economic impact 

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 § 50.109, 

‘‘Backfitting,’’ and apply to CPs and OLs 

under part 50. Issue finality provisions 

(analogous to the backfitting provisions 

in § 50.109) for approvals under part 52 

are located in various provisions of part 

52. The NRC Management Directive 8.4, 

‘‘Management of Backfitting, Forward 

Fitting, Issue Finality, and Information 

Requests,’’ describes the Commission’s 

policies on backfitting and issue 

finality. 

This proposed rule would provide a 

regulatory scheme for entities to apply 

for approvals under part 53. The part 50 

backfitting provisions and part 52 issue 

finality provisions apply to actions 

taken by the NRC under part 50 or part 

52, respectively, or actions taken by the 

NRC under other parts of 10 CFR 

chapter I that, for holders of certain 

approvals under part 50 or part 52, 

inextricably affect their activities 

regulated under part 50 or part 52. 

Issuance and implementation of 

proposed part 53 would not constitute 

actions taken under part 50 or part 52. 

Also, proposed part 53 would not allow 

an applicant to reference approvals 

issued under part 50 or part 52. 

Therefore, the issuance and 

implementation of proposed part 53 

would not affect part 50 or part 52 

entities’ activities regulated under part 

50 or part 52. Therefore, the addition of 

part 53 through this proposed rule 

would not be within the scope of the 

part 50 backfitting and part 52 issue 

finality provisions. 

The NRC also proposes conforming 

changes to parts 1, 2, 10, 11, 19, 20, 21, 

25, 26, 30, 40, 50, 51, 70, 72, 73, 74, 75, 

95, 140, 150, 170, and 171 to reflect the 

addition of part 53. These changes 

would not meet the definition of 

‘‘backfitting’’ in § 50.109 or § 70.76, 

‘‘Backfitting,’’ because the proposed 

changes would not modify or add to the 

systems, structures, components, or 


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design of a facility or to the procedures 

or organization required to operate a 

facility under part 50 or 70. These 

changes would not meet the definition 

of ‘‘backfitting’’ in § 72.62, 

‘‘Backfitting,’’ because the proposed 

changes would not add, eliminate, or 

modify the SSCs of an independent 

spent fuel storage installation (ISFSI) or 

the procedures or organization required 

to operate an ISFSI. These proposed 

changes would not inextricably affect 

activities regulated under parts 50, 52, 

70, or 72. Therefore, the proposed 

changes to parts 1, 2, 10, 11, 19, 20, 21, 

25, 26, 30, 40, 50, 51, 70, 72, 73, 74, 75, 

95, 140, 150, 170, and 171 would not 

constitute backfitting under parts 50, 70, 

or 72 or affect the issue finality of an 

approval under part 52. 

The NRC is issuing 10 draft guidance 

documents that, if issued as final 

guidance documents, would provide 

guidance on the methods acceptable to 

the NRC for complying with aspects of 

this proposed rule. These documents 

would not apply to holders of approvals 

issued under part 50 or part 52. Further, 

as discussed in the guidance 

documents, applicants and licensees 

would not be required to comply with 

the positions set forth in the guidance. 

Therefore, issuance of the guidance 

documents as final guidance would not 

constitute backfitting under part 50 or 

affect the issue finality of any approval 

issued under part 52. 

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. 


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The form number if applicable: NRC 

Forms 361S, 366, 366A, 366B, 893, and 

894. 

How often the collection is required or 

requested: Once, on occasion, every 30 

days, biannually, annually, biennially, 

every four years, every five years, every 

ten years. 

Who will be required or asked to 

respond: Part 53 commercial nuclear 

plant licensees and license applicants 

for commercial nuclear plants to be 

licensed under part 53. 

An estimate of the number of annual 

responses: 15 (2 responses for Part 26, 

11 responses for Part 53, 2 responses for 

Part 50 and 0 responses for Part 73 and 

NRC Forms 361S, 366, 366A, 366B, 893, 

and 894) 

The estimated number of annual 

respondents: 2 (2 respondents for Part 

26, 2 respondents for Part 53, 2 

respondents for Part 50 and 0 

respondents for Part 73 and NRC Forms 

361S, 366, 366A, 366B, 893, and 894) 

An estimate of the total number of 

hours needed annually to comply with 

the information collection requirement 

or request: 230,244 hours. (656 hours for 

Part 26, 220,801 hours for Part 53, 8,767 

hours for Part 50 and 0 hours for Part 

73 and NRC Forms 361S, 366, 366A, 

366B, 893, and 894) 

Abstract: The NRC is proposing to 

establish an optional technologyinclusive regulatory framework for use 

by applicants for new commercial 

nuclear plant designs. The regulatory 

requirements developed in this 

rulemaking would use methods of 

evaluation, including risk-informed and 

performance-based methods, that are 

flexible and practicable for application 

to a variety of new reactor technologies. 

The NRC’s goals in amending these 

regulations are to continue to provide 

reasonable assurance of adequate 

protection of public health and safety 

and the common defense and security at 

reactor sites at which new nuclear 

reactor designs are deployed to at least 

the same degree of protection as 

required for current-generation LWRs; 

protect health and minimize danger to 

life or property to at least the same 

degree of protection as required for 

current-generation LWRs; provide 

greater operational flexibilities where 

supported by enhanced margins of 

safety that may be provided in new 

nuclear designs; and promote regulatory 

stability, predictability, and clarity. 

The proposed rule covers diverse 

topics, which result in recordkeeping 

and reporting requirements related to 

contents of applications, plant design 

and analysis, siting, construction and 

manufacturing, licensing-basis 

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-forduty program performance (NRC Form 

894). Finally, a new version of NRC 

Form 361 (NRC Form 361S) would be 

created for use by part 53 licensees, 

covering an equivalent scope as the 

requirements in 10 CFR 50.72, but 

without LWR-specific terminology to 

ensure technology inclusiveness. 

The NRC is seeking public comment 

on the potential impact of the 

information collections contained in 

this proposed rule and on the following 

issues: 

1. Is the proposed information 

collection necessary for the proper 

performance of the functions of the 

NRC, including whether the information 

will have practical utility? Please 

explain your response. 

2. Is the estimate of the burden of the 

proposed information collection 

accurate? Please explain your response. 

3. Is there a way to enhance the 

quality, utility, and clarity of the 

information to be collected? Please 

explain your response. 

4. How can the burden of the 

proposed information collection on 

respondents be minimized, including 

the use of automated collection 

techniques or other forms of information 

technology? Please explain your 

response. 

The OMB clearance documents and 

proposed rule is available as indicated 

under the ‘‘Availability of Documents’’ 

section in this document or may be 

viewed free of charge by contacting the 

NRC’s PDR reference staff at 1–800– 

397–4209, at 301–415–4737, or by email 

to PDR.resource@nrc.gov. You may 

obtain information and comment 

submissions related to the OMB 

clearance package by searching on 

http://www.regulations.gov under 

Docket ID NRC–2019–0062. 

You may submit comments on any 

aspect of these proposed information 

collections, including suggestions for 

reducing the burden and on the above 

issues, by the following methods: 

• Federal Rulemaking Website: Go to 

http://www.regulations.gov and search 

for Docket ID NRC–2019–0062. 

• Mail comments to: FOIA, Library, 

and Information Collections Branch, 

Office of the Chief Information Officer, 

Mail Stop: T6–A10M, U.S. Nuclear 

Regulatory Commission, Washington, 

DC 20555–0001 or by email to 

Infocollects.Resource@nrc.gov or to the 

OMB reviewer at: OMB Office of 

Information and Regulatory Affairs 

(3150–XXXX, 3150–0002, –0104, –0146, 

–0238), Attn: Desk Officer for the 

Nuclear Regulatory Commission, 725 

17th Street NW, Washington, DC 20503. 

Submit comments by December 2, 

2024. Comments received after this date 

will be considered if it is practical to do 

so, but the NRC staff is able to ensure 

consideration only for comments 

received on or before this date. 

Public Protection Notification 

The NRC may not conduct or sponsor, 

and a person is not required to respond 

to, a collection of information unless the 

document requesting or requiring the 

collection displays a currently valid 

OMB control number. 

XV. Criminal Penalties 

For the purposes of Section 223 of the 

Act, the NRC is issuing this proposed 

rule that would add a new part 53 and 

amend parts 26 and 73 under one or 

more of Sections 161b, 161i, or 161o of 

the Act, except as noted in proposed 

§ 53.9010(b) and § 26.825(b). Willful 

violations of the part 53 and part 26 

regulations not listed in proposed 

§ 53.9010(b) and § 26.825(b) would be 

subject to criminal enforcement. 

Criminal penalties as they apply to 

regulations in part 53 would be 

discussed in § 53.9010. 

XVI. Voluntary Consensus Standards 

The NTTAA requires that Federal 

agencies use technical standards that are 

developed or adopted by voluntary 

consensus standards bodies unless the 

use of such a standard is inconsistent 

with applicable law or otherwise 

impractical. In this proposed rule, the 

NRC would revise regulations by adding 

a risk-informed, technology-inclusive 

regulatory framework for commercial 

advanced nuclear reactors. This action 

does not constitute the establishment of 

a standard that contains generally 

applicable requirements. 

XVII. Availability of Guidance 

As discussed in section II, 

Background, of this document, the 

NRC’s development of proposed part 53 


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built upon recent and ongoing activities 

such as those described in SECY–19– 

0117. Because a number of those 

activities are ongoing to support new 

reactor applications under the existing 

regulatory framework of 10 CFR parts 50 

and 52, the NRC staff identified in its 

response to SRM–SECY–20–0032 that 

the timing of guidance document 

development to support the part 53 

rulemaking was a key risk and 

uncertainty to publishing the final part 

53 rule. To mitigate this risk, the NRC 

engaged external stakeholders to ensure 

a common prioritization of the 

development of these guidance 

documents and to work diligently on 

those that would be needed to support 

this rulemaking, forthcoming 

applications, or broader efforts such as 

the Advanced Reactor Demonstration 

Program being sponsored by the DOE. 

The NRC also recognizes that guidance 

development to support part 53 and 

advanced reactors will continue as the 

industry and NRC learn lessons from 

licensing reviews and operating 

experience. Therefore, the NRC 

categorized guidance supporting the 

part 53 rulemaking into three categories: 

(1) guidance issued or under 

development to support applications 

under the existing regulatory 

framework; (2) implementing guidance 

for part 53-specific proposed rule 

language; and (3) future guidance 

activities that would need to be 

completed after the part 53 proposed 

rule is published for public comment. 

(1) Hundreds of guidance documents 

exist for the current fleet of operating 

reactors. While some of the guidance is 

specific to LWR technologies, other 

guidance is technology inclusive in 

nature and should be considered, as 

appropriate, in the development of all 

licensing applications and NRC reviews. 

In addition, the NRC has undertaken 

efforts to incorporate or reference the 

most relevant guidance in its efforts to 

develop additional guidance for future 

advanced reactors. The NRC has issued 

the following guidance to support 

licensing reviews of advanced reactors 

under the existing regulatory framework 

that will continue to inform applicant 

development and NRC reviews under 

parts 50 and 52. Conforming changes to 

these guidance documents would be 

needed to ensure they are applicable 

under part 53. The NRC will issue 

revisions or part 53-related companions 

to these guidance documents for public 

comment after the publication of this 

proposed rule and then finalize and 

issue the guidance documents with or 

after the final part 53 rule. 

• RG 1.233, ‘‘Guidance for a 

Technology-Inclusive, Risk-Informed, 

and Performance-Based Methodology 

to Inform the Licensing Basis and 

Content of Applications for Licenses, 

Certifications, and Approvals for NonLight-Water Reactors’’ 

• RG 1.247 for trial use, ‘‘Acceptability 

of Probabilistic Risk Assessment 

Results for Non-Light-Water Reactor 

Risk-Informed Activities’’ 

• NUREG–2246, ‘‘Fuel Qualification for 

Advanced Reactors’’ 

• RG 1.87, Revision 2, ‘‘Acceptability of 

ASME Code, Section III, Division 5, 

‘‘High Temperature Reactors’’ 

• RG 1.246, ‘‘Acceptability of ASME 

Code, Section XI, Division 2, 

‘Requirements for Reliability And 

Integrity Management (RIM) Programs 

for Nuclear Power Plants,’ for NonLight Water Reactors’’ 

Also, the NRC continues to develop 

additional guidance to support licensing 

reviews of advanced reactors under the 

existing regulatory framework. Some of 

these guidance documents have been 

issued and others will be issued before 

the finalization of part 53 to support 

near-term applicants and NRC reviews. 

For example, the NRC has been and 

continues to be engaged with the DOE 

and industry to develop content of 

application guidance and other 

regulatory guidance for advanced 

reactors to support applications and 

subsequent operations under the 

existing regulatory framework. These 

guidance documents, such as the 

industry-led Technology-Inclusive 

Content of Application Project guidance 

found in NEI 21–07, Revision 1, and the 

NRC-led Advanced Reactor Content of 

Application Project (ARCAP) interim 

staff guidance (ISG) documents and 

NRC regulatory guidance endorsing NEI 

21–07, Revision 1, will support 

developers in preparing advanced 

reactor applications. These guidance 

documents provide an overview of the 

information that should be included in 

an advanced reactor application, a 

review roadmap for the NRC with the 

principal purpose of ensuring 

consistency, quality, and uniformity of 

NRC reviews, and a well-defined base 

from which the NRC can evaluate 

proposed changes in the scope and 

requirements of reviews. While specific 

sections of the information are primarily 

aligned with the LMP methodology, as 

endorsed in RG 1.233, as one acceptable 

process for applicants to use when 

developing portions of an application, 

the concepts and general information 

may be used to inform the review of an 

application submitted using other 

traditional licensing approach 

methodologies (as applicable). Other 

sections of the information are generally 

applicable and independent of the 

methodology used to develop an 

advanced reactor application. The 

ARCAP ISGs provide references to 

numerous regulatory guidance 

documents that should be considered by 

both applicants and the NRC in 

developing and reviewing, respectively, 

advanced reactor applications. The NRC 

has issued the following documents 

separately from this proposed rule. The 

NRC may issue other, related guidance 

documents with or after the final part 53 

rule. 

• RG 1.253, ‘‘Guidance for a Technology 

Inclusive Content of Application 

Methodology to Inform the Licensing 

Basis and Content of Applications for 

Licenses, Certifications, and 

Approvals for Non-Light-Water 

Reactors’’ 

• DANU–ISG–2022–01, ‘‘Advanced 

Reactor Content of Application 

Project, ‘Review of Risk-Informed, 

Technology-Inclusive Advanced 

Reactor Applications—Roadmap’ ’’ 

• DANU–ISG–2022–02, ‘‘Advanced 

Reactor Content of Application 

Project Chapter 2, ‘Site Information’ ’’ 

• DANU–ISG–2022–03, ‘‘Advanced 

Reactor Content of Application 

Project Chapter 9, ‘Control of Routine 

Plant Radioactive Effluents, Plant 

Contamination and Solid Waste’ ’’ 

• DANU–ISG–2022–04, ‘‘Advanced 

Reactor Content of Application 

Project Chapter 10, ‘Control of 

Occupational Dose’ ’’ 

• DANU–ISG–2022–05, ‘‘Advanced 

Reactor Content of Application 

Project Chapter 11, ‘Organization and 

Human-System Considerations’ ’’ 

• DANU–ISG–2022–06, ‘‘Advanced 

Reactor Content of Application 

Project Chapter 12, ‘Post-Construction 

Inspection, Testing, and Analysis 

Program’ ’’ 

• DANU–ISG–2022–07, ‘‘Advanced 

Reactor Content of Application 

Project, ‘Risk-Informed Inservice 

Inspection/Inservice Testing’ ’’ 

• DANU–ISG–2022–08, ‘‘Advanced 

Reactor Content of Application 

Project, ‘Risk-Informed Technical 

Specifications’ ’’ 

• DANU–ISG–2022–09, ‘‘Advanced 

Reactor Content of Application 

Project, ‘Risk-Informed, PerformanceBased Fire Protection Program (for 

Operations)’ ’’ 

• RG 1.242, ‘‘Performance-Based 

Emergency Preparedness for Small 

Modular Reactors, Non-Light-Water 

Reactors, and Non-Power Production 

or Utilization Facilities’’ 

• RG 4.7, ‘‘General Site Suitability 

Criteria for Nuclear Power Stations’’ 


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Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules 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 § 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 § 73.120, 

‘‘Access authorization program for 

commercial nuclear plants,’’ related to 

an AA program. This document 

provides guidance and would be one 

NRC-approved method (not the only 

method) for meeting regulatory 

requirements for part 53. The proposed 

language in § 73.120 would provide 

flexibility through availability of the use 

of an alternate approach, commensurate 

with risk and consequence to public 

health and safety, for part 53 applicants 

who demonstrate in an analysis that the 

offsite consequences satisfy the criterion 

defined in proposed § 53.860(a)(2)(i). 

• DG–5075, ‘‘Establishing Cybersecurity 

Programs for Commercial Nuclear 

Plants Licensed Under 10 CFR part 

53’’ 

This DG describes an approach the 

NRC staff deems acceptable for 

complying with the Commission’s 

proposed regulations for establishing, 

implementing, and maintaining a 

cybersecurity program at commercial 

nuclear plants that would be licensed 

under part 53. This guidance provides 

an approach for meeting the 

requirements of proposed § 73.110, 

‘‘Technology-inclusive requirements for 

protection of digital computer and 

communication systems and networks.’’ 

• DG–5076, ‘‘Guidance for Technology 

Inclusive Requirements for Physical 

Protection of Licensed Activities at 

Commercial Nuclear Plants’’ 

This DG describes methods and 

approaches that the NRC staff considers 

acceptable for meeting the proposed 

physical security requirements of part 

53 and § 73.100. The guidance is 

intended to provide methods and 

considerations for complying with 

§ 53.440(f) safety and security design 

process considerations, determining 

eligibility for meeting the performance 

criterion in § 53.860 to relieve the 

applicant from the applicable 

requirements to defend against 

radiological sabotage outlined in § 73.55 

or § 73.100, and (if the required analysis 

for eligibility is not satisfied) applying 

the physical security requirements of 

§ 73.100 as an alternative pathway from 

§ 73.55 for protection against 

radiological sabotage. 

• DG–5078, ‘‘Fatigue Management for 

Nuclear Power Plant Personnel at 

Commercial Nuclear Plants Licensed 

Under 10 CFR part 53’’ 

This DG describes proposed methods 

that the NRC staff considers acceptable 

for addressing certain aspects of FFD 

programs that would be established at 

commercial nuclear facilities licensed 

under part 53. This guidance, in 

conjunction with the existing RG 5.73, 

‘‘Fatigue Management for Nuclear Plant 

Personnel,’’ would provide 

comprehensive guidance regarding 

acceptable methods for the development 

and implementation of licensee fatiguemanagement programs. 

The NRC is issuing for public 

comment the following draft ISG 

documents for the implementation of 

NRC staff review of applications under 

the proposed requirements in this 

rulemaking: 

• DRO–ISG–2023–01, ‘‘Operator 

Licensing Programs’’ 

This draft ISG provides guidance for 

the review of tailored operator licensing 

programs that are submitted for review 

consistent with the technical 

requirements of proposed § 53.730(g). 

This guidance primarily addresses the 

review of operator licensing 

examination processes to facilitate the 

ability of reviewers to assess whether a 

proposed approach to the testing of 

licensed operators and trainees reflects 

sound assessment testing practices that 

are suitable for the screening of 

competent licensed operators. 

Additionally, this ISG provides further 

review guidance in other areas such as 

licensed operator continuing training 

and proficiency programs. 

• DRO–ISG–2023–02, ‘‘Interim Staff 

Guidance Augmenting NUREG–1791, 

‘Guidance for Assessing Exemption 

Requests from the Nuclear Power 

Plant Licensed Operator Staffing 

Requirements Specified in 10 CFR 

50.54(m),’ for Licensing Commercial 

Nuclear Plants under 10 CFR part 53’’ 

This draft ISG provides guidance for 

the review of customized facility 

operator staffing plans that are 

submitted for review consistent with the 

technical requirements of proposed 

§ 53.730(f). This ISG is structured as a 

companion document to the existing 

NUREG–1791 and adapts the existing 

HFE-based methodologies of that 

document for use in the evaluation of 

staffing plans that would be submitted 

within the context of part 53 facilities. 

Additionally, this ISG provides further 

guidance to address other staffingrelated considerations, such as 

provisions for engineering expertise. 

• DRO–ISG–2023–03, ‘‘Development of 

Scalable Human Factors Engineering 

Review Plans’’ 


lotter on DSK11XQN23PROD with PROPOSALS2
87012 Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules 

This draft ISG applies to the HFE 

review of applications for OLs, COLs, 

DCs, and standard design approvals for 

commercial nuclear plants submitted 

under proposed part 53. The purpose of 

this ISG is to facilitate NRC 

understanding of an acceptable method 

for developing a scalable (i.e., 

application-specific) plan for the review 

of these applications for compliance 

with applicable HFE requirements. The 

ISG describes a process and provides 

implementation guidance for the NRC to 

tailor HFE review plans to each 

application to achieve an effective and 

efficient review. 

(3) The NRC has identified future 

guidance activities that would need to 

be completed after the part 53 proposed 

rule is published for public comment to 

support advanced reactor applications 

and NRC reviews. For example, the NRC 

recognizes that new guidance would be 

needed for the implementation of 

provisions in proposed § 53.620(d) and 

the associated licensing provisions in 

proposed subpart H that would allow 

and establish requirements for the 

loading of fuel into a manufactured 

reactor for subsequent transport to and 

use at a commercial nuclear plant that 

will operate the facility pursuant to a 

COL. The NRC has not yet initiated the 

development of guidance documents in 

this category but will engage 

stakeholders during the development of 

these documents to ensure common 

prioritization. In addition, the NRC 

works with standards development 

organizations, advanced reactor 

developers, DOE, and other stakeholders 

to identify and facilitate new consensus 

codes and standards needed for 

advanced reactor development. The 

NRC will continue its membership and 

participation on standards development 

committees and working groups to 

support standards for advanced reactor 

technologies, where appropriate. 

XVIII. Public Meeting 

The NRC will conduct a public 

meeting on this proposed rule for the 

purpose of describing the proposed rule 

and implementation guidance to the 

public and answering questions from 

the public on the proposed rule and 

implementation guidance. 

The NRC will publish a notice of the 

public meeting’s location, time, and 

agenda on the NRC’s public meeting 

website at least 10 calendar days before 

the meeting. Stakeholders should 

monitor the NRC’s public meeting 

website for information about the public 

meeting at: https://www.nrc.gov/publicinvolve/public-meetings/index.cfm. 

XIX. Availability of Documents 

The documents identified in the 

following table are available to 

interested persons through one or more 

of the following methods, as indicated. 

Document ADAMS accession No./Web link/ 

Federal Register Citation 

Proposed Rule Documents 

Federal Register Notification, ‘‘Proposed Rule: Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors,’’ October, 2024.ML24095A161. 

‘‘Draft Environmental Assessment for the Proposed Rule—Risk Informed, Technology-Inclusive 

Regulatory Framework for Advanced Reactors,’’ October, 2024.

ML24095A163. 

‘‘Draft Regulatory Analysis for the Proposed Rule: Risk Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors,’’ October, 2024.ML24095A166. 

Information Collection Documents 

Draft Supporting Statement for Information Collection Analysis—10 CFR Part 53 ....................... ML21162A109. 

Draft Supporting Statement for Information Collection Analysis—10 CFR Part 26 ....................... ML23030A400. 

Draft Supporting Statement for Information Collection Analysis—10 CFR Part 50 ....................... ML24220A036. 

Draft Supporting Statement for Information Collection Analysis—10 CFR Part 73 ....................... ML23030A576. 

Draft Supporting Statement for Information Collection Analysis—NRC Form 361S ...................... ML24220A034. 

Draft Supporting Statement for Information Collection Analysis—NRC Form 366 ........................ ML24220A035. 

Draft Supporting Statement for Information Collection Analysis—NRC Form 893 and 894 .......... ML24220A033. 

Proposed Rule—Part 26 Burden Tables for Risk Informed, Technology-Inclusive Regulatory 

Framework for Advanced Reactors.

ML24240A008. 

Proposed Rule—Part 50 Burden Tables for Risk Informed, Technology-Inclusive Regulatory 

Framework for Advanced Reactors.

ML24220A061. 

Proposed Rule—Part 53 Burden Tables for Risk Informed, Technology-Inclusive Regulatory 

Framework for Advanced Reactors.

ML24220A060. 

Proposed Rule—Part 73 Burden Tables for Risk Informed, Technology-Inclusive Regulatory 

Framework for Advanced Reactors.

ML24240A009. 

Draft NRC Form 361S, ‘‘Part 53 Plant Event Notification Worksheet’’ .......................................... ML23032A443. 

Draft NRC Form 366, ‘‘Licensee Event Report (LER)’’ .................................................................. ML23032A445. 

Draft NRC Form 366A, ‘‘Licensee Event Report (LER) Continuation Sheet’’ ................................ ML23032A447. 

Draft NRC Form 366B, ‘‘Licensee Event Report (LER) (Failure Continuation)’’ ............................ ML23032A454. 

Draft NRC Form 893, ‘‘10 CFR Part 26, Subpart M, Single FFD Policy Violation Form’’ ............. ML23032A435. 

Draft NRC Form 894, ‘‘10 CFR Part 26, Subpart M, Annual Reporting Form for FFD Performance Information’’.ML23032A439. 

Draft Regulatory Guidance Documents 

DG–1413, ‘‘Technology-Inclusive Identification Of Licensing Events For Commercial Nuclear 

Plants,’’ October, 2024.

ML22257A173. 

DG–5073, ‘‘Fitness-For-Duty Programs For Commercial Nuclear Plants And Manufacturing Facilities Licensed Under 10 CFR Part 53,’’ October, 2024.ML22200A037. 

DG–5074, ‘‘Access Authorization Program for Commercial Nuclear Plants,’’ October, 2024 ....... ML22199A246. 

DG–5075, ‘‘Establishing Cybersecurity Programs For Commercial Nuclear Plants Licensed 

Under 10 CFR Part 53,’’ October, 2024.

ML22199A257. 

DG–5076, ‘‘Guidance for Technology Inclusive Requirements for Physical Protection of Licensed Activities at Commercial Nuclear Plants,’’ October, 2024.ML22203A131. 


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Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules 87013 

Document ADAMS accession No./Web link/ 

Federal Register Citation 

DG–5078, ‘‘Fatigue Management For Nuclear Power Plant Personnel At Commercial Nuclear 

Plants Licensed Under 10 CFR Part 53,’’ October, 2024.

ML22264A109. 

Draft ISG Documents 

Draft ISG DRO–ISG–2023–01, ‘‘Operator Licensing Programs,’’ October, 2024 .......................... ML22266A066. 

Draft ISG DRO–ISG–2023–02, ‘‘Interim Staff Guidance Augmenting NUREG–1791, ‘Guidance 

for Assessing Exemption Requests from the Nuclear Power Plant Licensed Operator Staffing 

Requirements Specified in 10 CFR 50.54(m),’ for Licensing Commercial Nuclear Plants 

under 10 CFR Part 53,’’ October, 2024.

ML22266A068. 

Draft ISG DRO–ISG–2023–03, ‘‘Development of Scalable Human Factors Engineering Review 

Plans,’’ October, 2024.

ML22266A072. 

Other References 

American National Standards Institute/ANS–3.4–2013, ‘‘Medical Certification And Monitoring Of 

Personnel Requiring Operator Licenses For Nuclear Power Plants’’.

https://webstore.ansi.org/Standards/ANSI/ 

ansians2013. 

ASME/ANS RA–S–1.4–2021, ‘‘Probabilistic Risk Assessment Standard for Advanced Non-Light 

Water Reactor Nuclear Power Plants’’.

https://www.asme.org/codes-standards/findcodes-standards/probabilistic-risk-assessment-standard-for-advanced-non-light-waterreactor-nuclear-power-plants/2021/pdf. 

ASCE/SEI 43–19, ‘‘Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities’’.https://doi.org/10.1061/9780784415405. 

Federal Register notification—Final policy statement, ‘‘Use of Probabilistic Risk Assessment 

Methods in Nuclear Regulatory Activities; Final Policy Statement,’’ dated August 16, 1995.

60 FR 42622. 

Federal Register notification—Final rule, ‘‘Fitness-for-Duty Programs,’’ dated June 7, 1989 ...... 54 FR 24468. 

Federal Register notification—Final rule, ‘‘Fitness for Duty Programs,’’ dated March 31, 2008 .. 73 FR 16966. 

Federal Register notification—Final rule, ‘‘Licenses, Certifications, and Approvals for Nuclear 

Power Plants,’’ dated August 28, 2007.

72 FR 49352. 

Federal Register notification—Final rule, ‘‘Station Blackout,’’ dated June 21, 1988 .................... 53 FR 23203. 

Federal Register notification—Final rule, ‘‘Technical Specifications,’’ dated July 19, 1995 ......... 60 FR 36953, 36955. 

Federal Register notification—Guidance, ‘‘Mandatory Guidelines for Federal Workplace Drug 

Testing Programs,’’ dated January 23, 2017.

82 FR 7920. 

Federal Register notification—Guidance, ‘‘Mandatory Guidelines for Federal Workplace Drug 

Testing Programs—Oral/Fluid,’’ dated October 25, 2019.

84 FR 57554. 

Federal Register notification—Policy Statement, ‘‘Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants,’’ dated August 8, 1985.50 FR 32138. 

Federal Register notification—Policy Statement, ‘‘Safety Goals for the Operation of Nuclear 

Power Plants; Policy Statement; Correction and Republication,’’ dated August 21, 1986.

51 FR 30028. 

Federal Register notification—Policy Statement, ‘‘Tribal Policy Statement,’’ dated January 9, 

2017.

82 FR 2402. 

Federal Register notification—Policy Statement, ‘‘Policy Statement on the Regulation of Advanced Reactors,’’ dated October 14, 2008.73 FR 60612. 

Federal Register notification—Policy Statement, ‘‘Final Safety Culture Policy Statement,’’ 

dated June 14, 2011.

76 FR 34773. 

Federal Register notification—Proposed rule, ‘‘Emergency Preparedness for Small Modular 

Reactors and Other New Technologies,’’ dated May 12, 2020.

85 FR 28436. 

Federal Register notification—Proposed rule, ‘‘Regulatory Improvements for Production and 

Utilization Facilities Transitioning to Decommissioning,’’ dated March 3, 2022.

87 FR 12254. 

Federal Register notification—Public meeting, ‘‘Reporting Requirements for Nonemergency 

Events at Nuclear Power Plants,’’ dated November 29, 2021.

86 FR 67669. 

ICRP, Publication 2 ‘‘Permissible dose for internal radiation,’’ dated 1960 ................................... https://www.icrp.org/publication.asp?id=icrp%20

publication%202. 

ICRP, Publication 26 ‘‘Recommendations of the ICRP,’’ dated 1977 ............................................ https://www.icrp.org/publication.asp?id=

ICRP%20Publication%2026. 

ICRP, Publication 30 ‘‘Limits for Intakes of Radionuclides by Workers,’’ dated 1979 ................... https://www.icrp.org/publication.asp?id=

ICRP%20Publication%2030%20(Index). 

Letter to Chairman Hanson, NRC, ‘‘Final Letter on Draft 10 CFR Part 53 Rulemaking Language,’’ dated November 22, 2022.ML22319A104. 

Letter to Chairman Hanson, NRC, ‘‘Fourth Interim Letter on 10 CFR Part 53 Rulemaking Language,’’ dated August 2, 2022.ML22196A292. 

Letter to Chairman Hanson, NRC, ‘‘Preliminary Proposed Rule Language For 10 CFR Part 53, 

Regulation of Advanced Nuclear Reactors, Interim Report,’’ dated May 30, 2021.

ML21140A354. 

Letter to Chairman Hanson, NRC, ‘‘Preliminary Rule Language For 10 CFR Part 53, Subpart F, 

‘Requirements for Operations,’ Interim Report,’’ dated February 17, 2022.

ML22040A361. 

Letter to Chairman Rempe, ACRS, ‘‘Response to the Advisory Committee on Reactor Safeguards, ‘Fourth Interim Letter on 10 CFR Part 53 Rulemaking Language,’’’ dated September 

30, 2022.

ML22249A073. 

Letter to Chairman Rempe, ACRS, ‘‘Response to the Advisory Committee on Reactor Safeguards Letter on Preliminary Rule Language for 10 CFR Part 53, Subpart F, ‘Requirements 

for Operations,’ Interim Report,’’ dated March 30, 2022.

ML22063A012. 

Letter to Chairman Sunseri, ACRS, ‘‘Part 53, Licensing and Regulation of Advanced Nuclear 

Reactors,’’ dated November 24, 2020.

ML20311A006. 


lotter on DSK11XQN23PROD with PROPOSALS2
87014 Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules 

Document ADAMS accession No./Web link/ 

Federal Register Citation 

Letter to Chairman Svinicki, NRC, ‘‘10 CFR Part 53, Licensing and Regulation of Advanced 

Nuclear Reactors,’’ dated October 21, 2020.

ML20295A647. 

Michigan v. EPA, 135 S. Ct. 2699 (2015) ......................................................................................

National Library of Medicine, National Institutes of Health, Workshop Summary, ‘‘The Evolution 

of Telehealth: Where Have We Been and Where Are We Going?,’’ dated November 2012.

https://www.ncbi.nlm.nih.gov/books/ 

NBK207141/. 

NEI 18–04, Rev. 1, ‘‘Risk-Informed Performance-Based Technology-Inclusive Guidance for 

Non-Light Water Reactors,’’ dated August 2019.

ML19241A472. 

NIA, ‘‘Clarifying ‘Major Portions’ of a Reactor Design in Support of a Standard Design Approval,’’ dated April 2017.https://www.nuclearinnovationalliance.org/clarifying-major-portions-reactor-design-supportstandard-design-approval. 

NRC, ‘‘A Regulatory Review Roadmap for Non-Light Water Reactors,’’ dated December 2017 .. ML17312B567. 

NRC, ‘‘Manufacturing License ML–1 for Production of Up to Eight Floating Nuclear Plants,’’ 

dated September 30, 1982.

ML20070J215. 

NRC, ‘‘Risk-Informed and Performance-Based Human-System Considerations for Advanced 

Reactors,’’ dated March 2021.

ML21069A003. 

NRC Form 890, ‘‘Single Positive Test Form’’ ................................................................................. ML22013B187. 

NRC Form 891, ‘‘Annual Reporting for Drug and Alcohol Tests’’ .................................................. ML22013B240. 

NRC From 892, ‘‘Annual Fatigue Reporting Form’’ ........................................................................ ML22013B250. 

NUREG–0654/FEMA–REP–1, Revision 2, ‘‘Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants,’’ 

dated December 2019.

ML19347D139. 

NUREG–0880, ‘‘Safety Goals for Nuclear Power Plant Operation,’’ dated May 1983 .................. ML071770230. 

NUREG–1530, Revision 1, ‘‘Reassessment of NRC’s Dollar Per Person-Rem Conversion Factor Policy, Final Report,’’ dated February 2022.ML22053A025. 

NUREG/BR–0058, Revision 5, ‘‘Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory 

Commission,’’ dated April 2017.

ML17100A480. 

NUREG/CR–5884, ‘‘Revised Analyses of Decommissioning for the Reference Pressurized 

Water Reactor Power Station,’’ dated November 1995.

ML14008A187. 

NUREG/CR–6187, Volume 1, ‘‘Revised Analyses of Decommissioning for the Reference Boiling Water Reactor Power Station,’’ dated July 1996.ML14008A186. 

OMB Circular No. A–119, ‘‘Federal Participation in the Development and Use of Voluntary Consensus Standards and in Conformity Assessment Activities,’’ dated February 19, 1998.https://obamawhitehouse.archives.gov/omb/circulars_a119_a119fr. 

PNNL, Technical Letter Report, ‘‘The Use of Electronic Communications to Perform Determinations of Fitness,’’ dated August 2017.ML18081A607. 

Pre-decisional DG, ‘‘Technology-Inclusive, Risk-Informed, and Performance-Based Methodology for Seismic Design of Commercial Nuclear Plants,’’ dated October 3, 2022.ML22276A149. 

Research Information Letter 2021–04, ‘‘Feasibility Study on a Potential Consequence-Based 

Seismic Design Approach for Nuclear Facilities,’’ dated April 2021.

ML21113A066. 

RG 1.110, Revision 1, ‘‘Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled 

Nuclear Power Reactors,’’ dated October 2013.

ML13241A052. 

RG 1.134, Revision 4, ‘‘Medical Assessment Of Licensed Operators Or Applicants For Operator Licenses At Nuclear Power Plants,’’ dated September 2014.ML14189A385. 

RG 1.174, ‘‘An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions 

on Plant-Specific Changes to the Licensing Basis,’’ Revision 3, dated January 2018.

ML17317A256. 

RG 1.208, ‘‘A Performance-Based Approach to Define the Site-Specific Earthquake Ground 

Motion,’’ dated March 2007.

ML070310619. 

RG 1.232, ‘‘Guidance for Developing Principal Design Criteria for Non-Light-Water Reactors,’’ 

Revision 0, dated April 2018.

ML17325A611. 

RG 1.233, Revision 0, ‘‘Guidance for a Technology-Inclusive, Risk-Informed, and PerformanceBased Methodology to Inform the Licensing Basis and Content of Applications for Licenses, 

Certifications, and Approvals for Non-Light-Water Reactors,’’ dated June 2020.

ML20091L698. 

RG 1.247, ‘‘Acceptability of Probabilistic Risk Assessment Results for Non-Light-Water Reactor 

Risk-Informed Activities,’’ issued March 2022 for trial use.

ML21235A008. 

RG 5.73, ‘‘Fatigue Management for Nuclear Power Plant Personnel,’’ dated March 20, 2009 .... ML083450028. 

RG 5.77, ‘‘Insider Mitigation Program,’’ Revision 1, dated September 08, 2022 ........................... ML16342B024. 

RG 5.81, ‘‘Target Set Identification and Development for Nuclear Power Reactors,’’ Revision 1, 

dated December 2019 (non-public).

ML13151A355. 

SECY–18–0096, ‘‘Functional Containment Performance Criteria For Non-Light-Water-Reactors,’’ dated September 28, 2018.ML18115A157. 

SECY–19–0117, ‘‘Technology-Inclusive, Risk-Informed, and Performance-Based Methodology 

to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and 

Approvals for Non-Light-Water Reactors,’’ dated December 2019.

ML18311A264 (package). 

SECY–20–0032, ‘‘Rulemaking Plan on ‘Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors (RIN–3150–AK31; NRC–2019–0062,’’’ dated April 13, 2020.ML19340A056. 

SECY–20–0070, ‘‘(Redacted) Technical Evaluation of the Security Bounding Time Concept for 

Operating Nuclear Power Plants,’’ dated November 8, 2021.

ML20126G265 (package). 

SECY–22–0072, ‘‘Proposed Rule: Alternative Physical Security Requirements for Advanced 

Reactors (RIN 3150–AK19),’’ dated August 2, 2022.

ML21334A003 (package). 

SECY–83–293, ‘‘Amendments to 10 CFR 50 Related to Anticipated Transients Without Scram 

(ATWS) Events,’’ dated July 19, 1983.

ML21278A823 (non-public); ML21278A994 

(non-public). 

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.

ML040210725. 


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Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules 87015 

Document ADAMS accession No./Web link/ 

Federal Register Citation 

SRM–SECY–10–0121, ‘‘Modifying the Risk-Informed Regulatory Guidance for New Reactors,’’ 

dated March 2, 2011.

ML110610166. 

SRM–SECY–17–0100, ‘‘Security Baseline Inspection Program Assessment Results and Recommendations for Program Efficiencies,’’ dated October 8, 2018.ML18283A072. 

SRM–SECY–20–0032, ‘‘Rulemaking Plan on ‘Risk-Informed, Technology-Inclusive Regulatory 

Framework for Advanced Reactors (RIN–3150–AK31; NRC–2019–0062),’’’ dated October 2, 

2020.

ML20276A293. 

SRM–SECY–20–0045, ‘‘Population Related Siting Considerations for Advanced Reactors,’’ 

dated July 30, 2022.

ML22194A885. 

SRM–SECY–98–144, ‘‘Staff Requirements—SECY–98–144—White Paper on Risk-Informed 

and Performance-Based Regulations,’’ dated February 24, 1999.

ML003753593. 

SECY–23–0021, ‘‘Proposed Rule: Risk-Informed, Technology-Inclusive Regulatory Framework 

for Advanced Reactors (RIN 3150–AK31),’’ March 1, 2023.

ML21162A095. 

SECY–23–0021, Enclosure 1, ‘‘Draft Federal Register Notification’’ ............................................ ML21162A102. 

SECY–23–0021, Enclosure 2, ‘‘Draft Environmental Assessment for the Proposed Rule—Risk 

Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors’’.

ML21162A104. 

SECY–23–0021, Enclosure 3, ‘‘Draft Regulatory Analysis for the Proposed Rule: Risk Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors’’.ML21165A112. 

SECY–23–0021, Enclosure 4, ‘‘Alternative Approaches Considered for Selected Topics During 

the Development of 10 CFR Part 53’’.

ML22244A001. 

SECY–23–0021, Enclosure 5, ‘‘Estimated Resources for The Risk-Informed, Technology-Inclusive Regulatory Framework For Advanced Reactors Rulemaking’’.ML22304A099 (non-public). 

Staff Requirements—SECY–23–0021, ‘‘Proposed Rule: Risk-Informed, Technology-Inclusive 

Regulatory Framework for Advanced Reactors (RIN 3150–AK31),’’ March 4, 2024.

ML24064A047 (package). 

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 


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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 

■ 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). 

§ 1.43 [Amended] 

■ 2. In § 1.43, in paragraph (a)(2) remove 

the cross reference ‘‘10 CFR parts 50, 52, 

and 54’’ and add in its place the cross 

reference ‘‘10 CFR parts 50, 52, 53, and 

54’’. 

PART 2—AGENCY RULES OF 

PRACTICE AND PROCEDURE 

■ 3. The authority citation for part 2 

continues to read as follows: 

Authority: Atomic Energy Act of 1954, 

secs. 29, 53, 62, 63, 81, 102, 103, 104, 105, 

161, 181, 182, 183, 184, 186, 189, 191, 234 

(42 U.S.C. 2039, 2073, 2092, 2093, 2111, 

2132, 2133, 2134, 2135, 2201, 2231, 2232, 

2233, 2234, 2236, 2239, 2241, 2282); Energy 

Reorganization Act of 1974, secs. 201, 206 

(42 U.S.C. 5841, 5846); Nuclear Waste Policy 

Act of 1982, secs. 114(f), 134, 135, 141 

(42 U.S.C. 10134(f), 10154, 10155, 10161); 

Administrative Procedure Act (5 U.S.C. 552, 

553, 554, 557, 558); National Environmental 

Policy Act of 1969 (42 U.S.C. 4332); 44 U.S.C. 

3504 note. Section 2.205(j) also issued under 

28 U.S.C. 2461 note. 

§ 2.1 [Amended] 

■ 4. In § 2.1, in paragraph (e) remove the 

phrase ‘‘part 52’’ and add in its place 

the phrase ‘‘part 52 or part 53’’. 

■ 5. In § 2.4, revise the definitions for 

‘‘Contested proceeding’’ and ‘‘Facility’’ 

to read as follows: 

§ 2.4 Definitions. 

* * * * * 

Contested proceeding means— 

(1) A proceeding in which there is a 

controversy between the NRC staff and 

the applicant for a license or permit 

concerning the issuance of the license or 

permit or any of the terms or conditions 

thereof; 

(2) A proceeding in which the NRC is 

imposing a civil penalty or other 

enforcement action, and the subject of 

the civil penalty or enforcement action 

is an applicant for or holder of a license 

or permit, or is or was an applicant for 

or holder of a license or permit, or is or 

was an applicant for a standard design 

certification under part 52 or part 53 of 

this chapter; and 

(3) A proceeding in which a petition 

for leave to intervene in opposition to 

an application for a license or permit 

has been granted or is pending before 

the Commission. 

* * * * * 

Facility means production facility or a 

utilization facility as defined in §§ 50.2 

and 53.020 of this chapter. 

* * * * * 


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§ 2.100 [Amended] 

■ 6. In § 2.100, remove the phrase 

‘‘subpart E of part 52’’ and add in its 

place the phrase ‘‘subpart E of part 52 

or subpart H of part 53’’. 

■ 7. In § 2.101, revise paragraphs 

(a)(3)(i), (a)(5), (a)(9) introductory text 

and paragraph (a)(9)(i) to read as 

follows: 

§ 2.101 Filing of application. 

(a) * * * 

(3) * * * 

(i) Submit to the Director, Office of 

Nuclear Reactor Regulation, or Director, 

Office of Nuclear Material Safety and 

Safeguards, as appropriate, such 

additional copies as the regulations in 

part 50, subpart A of part 51, and part 

53 of this chapter require; 

* * * * * 

(5) An applicant for a construction 

permit under parts 50 or 53 of this 

chapter or a combined license under 

parts 52 or 53 of this chapter for a 

production or utilization facility which 

is subject to § 51.20(b) of this chapter, 

and is of the type specified in 

§ 50.21(b)(2) or (b)(3); or § 50.22; or part 

53, as applicable, of this chapter, or is 

a testing facility, may submit the 

information required of applicants by 

parts 50, 52, or 53 of this chapter in two 

parts. One part shall be accompanied by 

the information required by § 50.30(f) of 

this chapter, § 52.80(b) of this chapter, 

or § 53.1100(f) of this chapter, as 

applicable. The other part shall include 

any information required by § 50.34(a) 

and, if applicable, § 50.34a of this 

chapter; or §§ 52.79 and 52.80(a) of this 

chapter; or §§ 53.1109, 53.1306, 

53.1309, and 53.1312 of this chapter; or 

§§ 53.1109, 53.1413, 53.1416, and 

53.1419 of this chapter, as applicable. 

One part may precede or follow other 

parts by no longer than 6 months. If it 

is determined that either of the parts as 

described above is incomplete and not 

acceptable for processing, the Director, 

Office of Nuclear Reactor Regulation, or 

Director, Office of Nuclear Material 

Safety and Safeguards, as appropriate, 

will inform the applicant of this 

determination and the respects in which 

the document is deficient. Such a 

determination of completeness will 

generally be made within a period of 30 

days. Whichever part is filed first shall 

also include the fee required by 

§ 50.30(e) or § 53.1100(e) and § 170.21 of 

this chapter and the information 

required by §§ 50.33, 50.34(a)(1), and 

52.79(a)(1) of this chapter; or 

§§ 53.1109, 53.1309, and 53.1416 of this 

chapter, as applicable, and § 50.37 or 

§ 53.1115, as applicable, of this chapter. 

The Director, Office of Nuclear Reactor 

Regulation, or Director, Office of 

Nuclear Material Safety and Safeguards, 

as appropriate, will accept for docketing 

an application for a construction permit 

under part 50 or part 53 of this chapter 

or a combined license under parts 52 or 

53 of this chapter for a production or 

utilization facility which is subject to 

§ 51.20(b) of this chapter, and is of the 

type specified in § 50.21(b)(2) or (b)(3), 

or § 50.22, or part 53, as applicable, of 

this chapter or is a testing facility where 

one part of the application as described 

above is complete and conforms to the 

requirements of part 50 of this chapter. 

The additional parts will be docketed 

upon a determination by the Director, 

Office of Nuclear Reactor Regulation, or 

Director, Office of Nuclear Material 

Safety and Safeguards, as appropriate, 

that it is complete. 

* * * * * 

(9) An applicant for a construction 

permit for a utilization facility which is 

subject to § 51.20(b) of this chapter and 

is of the type specified in § 50.21(b)(2) 

or (b)(3), or § 50.22, or part 53 of this 

chapter, an applicant for or holder of an 

early site permit under part 52 or part 

53 of this chapter, or an applicant for a 

combined license under parts 52 or 53 

of this chapter, who seeks to conduct 

the activities authorized under 

§ 50.10(d) or § 53.1130 of this chapter 

may submit a complete application 

under paragraphs (a)(1) through (a)(4) of 

this section which includes the 

information required by § 50.10(d) or 

§ 53.1130 of this chapter. Alternatively, 

the applicant (other than an applicant 

for or holder of an early site permit) may 

submit its application in two parts: 

(i) Part one must include the 

information required by § 50.33(a) 

through (f) or § 53.1109(a) through (e) 

and § 53.1306 of this chapter, and the 

information required by § 50.10(d)(2) 

and (d)(3) or § 53.1130(a)(2) and (a)(3) of 

this chapter, as applicable. 

* * * * * 

■ 8. In § 2.104, revise paragraph (a) to 

read as follows: 

§ 2.104 Notice of hearing. 

(a) In the case of an application on 

which a hearing is required by the Act 

or this chapter, or in which the 

Commission finds that a hearing is 

required in the public interest, the 

Secretary will issue a notice of hearing 

to be published in the Federal Register. 

The notice must be published at least 15 

days, and in the case of an application 

concerning a limited work 

authorization, construction permit, early 

site permit, or combined license for a 

facility of the type described in 

§ 50.21(b) or 50.22, or subpart H of part 

53 of this chapter, as applicable, or a 

testing facility, at least 30 days, before 

the date set for hearing in the notice.1

In addition, in the case of an application 

for a limited work authorization, 

construction permit, early site permit, or 

combined license for a facility of the 

type described in § 50.22 or subpart H 

of part 53 of this chapter, as applicable, 

or a testing facility, the notice must be 

issued as soon as practicable after the 

NRC has docketed the application. If the 

Commission decides, under 

§ 2.101(a)(2), to determine the 

acceptability of the application based on 

its technical adequacy as well as 

completeness, the notice must be issued 

as soon as practicable after the 

application has been tendered. 

* * * * * 1 If the notice of hearing concerning an 

application for a limited work authorization, 

construction permit, early site permit, or 

combined license for a facility of the type 

described in § 50.21(b) or § 50.22, or subpart 

H of part 53 of this chapter, as applicable, or 

a testing facility, does not specify the time 

and place of initial hearing, a subsequent 

notice will be published in the Federal 

Register which will provide at least 30-day 

notice of the time and place of that hearing. 

After this notice is given, the presiding 

officer may reschedule the commencement of 

the initial hearing for a later date or 

reconvene a recessed hearing without again 

providing at least 30-day notice. 

■ 9. In § 2.105, revise paragraph (a) 

introductory text and paragraphs (a)(4), 

(a)(10), (a)(12), (a)(13), (b)(3) 

introductory text, (b)(3)(i), (ii), and (iv) 

to read as follows: 

§ 2.105 Notice of proposed action. 

(a) If a hearing is not required by the 

Act or this chapter, and if the 

Commission has not found that a 

hearing is in the public interest, it will, 

before acting thereon, publish in the 

Federal Register, as applicable, or on 

the NRC’s website, http://www.nrc.gov, 

or both, at the Commission’s discretion, 

either a notice of intended operation 

under § 52.103(a) or § 53.1452(a) of this 

chapter, as applicable, and a proposed 

finding that inspections, tests, analyses, 

and acceptance criteria for a combined 

license under subpart C of part 52 or 

under subpart H of part 53 of this 

chapter, have been or will be met, or a 

notice of proposed action with respect 

to an application for: 

* * * * * 

(4) An amendment to an operating 

license, combined license, or 

manufacturing license for a facility 

licensed under § 50.21(b) or § 50.22 or 

under subpart H of part 53 of this 

chapter, as applicable, or for a testing 

facility, as follows: 

(i) If the Commission determines 

under § 50.58 or § 53.1515 of this 


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chapter that the amendment involves no 

significant hazards consideration, 

though it will provide notice of 

opportunity for a hearing pursuant to 

this section, it may make the 

amendment immediately effective and 

grant a hearing thereafter; or 

(ii) If the Commission determines 

under §§ 50.58 and 50.91 or § 53.1515 of 

this chapter, as applicable, that an 

emergency situation exists or that 

exigent circumstances exist and that the 

amendment involves no significant 

hazards consideration, it will provide 

notice of opportunity for a hearing 

pursuant to § 2.106 (if a hearing is 

requested, it will be held after issuance 

of the amendment); 

* * * * * 

(10) In the case of an application for 

an operating license for a facility of a 

type described in § 50.21(b) or § 50.22, 

or part 53 of this chapter or a testing 

facility, a notice of opportunity for 

hearing shall be issued as soon as 

practicable after the application has 

been docketed; or 

* * * * * 

(12) An amendment to an early site 

permit issued under subpart A of part 

52, or under subpart H of part 53 of this 

chapter, as follows: 

(i) If the early site permit does not 

provide authority to conduct the 

activities allowed under § 50.10(e)(1) or 

§ 53.1130(b)(1) of this chapter, the 

amendment will involve no significant 

hazards consideration, and though the 

NRC will provide notice of opportunity 

for a hearing under this section, it may 

make the amendment immediately 

effective and grant a hearing thereafter; 

and 

(ii) If the early site permit provides 

authority to conduct the activities 

allowed under § 50.10(e)(1) or 

§ 53.1130(b)(1) of this chapter and the 

Commission determines under §§ 50.58 

and 50.91 or § 53.1515 of this chapter 

that an emergency situation exists or 

that exigent circumstances exist and 

that the amendment involves no 

significant hazards consideration, it will 

provide notice of opportunity for a 

hearing under § 2.106 of this chapter (if 

a hearing is requested, which will be 

held after issuance of the amendment). 

(13) A manufacturing license under 

subpart F of part 52 or subpart H of part 

53 of this chapter. 

(b) * * * 

(3) For a notice of intended operation 

under § 52.103(a) or § 53.1452(a) of this 

chapter, the following information: 

(i) The identification of the NRC 

action as making the finding required 

under § 52.103(g) or § 53.1452(g) of this 

chapter; 

(ii) The manner in which the licensee 

notifications under § 52.99(c) or 

§ 53.1449(c), of this chapter which are 

required to be made available by 

§ 52.99(e)(2) or § 53.1449(e)(2), of this 

chapter may be obtained and examined; 

* * * * * 

(iv) Any conditions, limitations, or 

restrictions to be placed on the license 

in connection with the finding under 

§ 52.103(g) or § 53.1452(g) of this 

chapter, and the expiration date or 

circumstances (if any) under which the 

conditions, limitations or restrictions 

will no longer apply. 

* * * * * 

■ 10. In § 2.106, revise paragraphs (a)(2), 

(a)(3), and (b)(2) introductory text to 

read as follows: 

§ 2.106 Notice of issuance. 

(a) * * * 

(2) An amendment of a license for a 

facility of the type described in 

§ 50.21(b) or § 50.22, or part 53 of this 

chapter, as applicable, or a testing 

facility, whether or not a notice of 

proposed action has been previously 

published; and 

(3) The finding under § 52.103(g) or 

§ 53.1452(g) of this chapter. 

(b) * * * 

(2) In the case of a finding under 

§ 52.103(g) or § 53.1452(g) of this 

chapter: 

* * * * * 

■ 11. In § 2.109, revise paragraphs (b), 

(c), and (d) to read as follows: 

§ 2.109 Effect of timely renewal 

application. 

* * * * * 

(b) If the licensee of a nuclear power 

plant licensed under § 50.21(b) or 

§ 50.22 or under subpart H of part 53 of 

this chapter files a sufficient application 

for renewal of either an operating 

license or a combined license at least 5 

years before the expiration of the 

existing license, the existing license will 

not be deemed to have expired until the 

application has been finally determined. 

(c) If the holder of an early site permit 

licensed under subpart A of part 52 or 

under subpart H of part 53 of this 

chapter, as applicable, files a sufficient 

application for renewal under § 52.29 or 

§ 53.1173 of this chapter, as applicable, 

at least 12 months before the expiration 

of the existing early site permit, the 

existing permit will not be deemed to 

have expired until the application has 

been finally determined. 

(d) If the licensee of a manufacturing 

license under subpart F of part 52, or 

under subpart H of part 53 of this 

chapter files a sufficient application for 

renewal under § 52.177 or § 53.1295 of 

this chapter at least 12 months before 

the expiration of the existing license, 

the existing license will not be deemed 

to have expired until the application has 

been finally determined. 

* * * * * 

■ 12. In § 2.110, revise paragraphs (a)(1) 

and (b) to read as follows: 

§ 2.110 Filing and administrative action on 

submittals for standard design approval or 

early review of site suitability issues. 

(a)(1) A submittal for a standard 

design approval under subpart E of part 

52 or under subpart H of part 53 of this 

chapter shall be subject to §§ 2.101(a) 

and 2.390 to the same extent as if it were 

an application for a permit or license. 

* * * * * 

(b) Upon initiation of review by the 

NRC staff of a submittal for an early 

review of site suitability issues under 

appendix Q to part 50 of this chapter, 

or for a standard design approval under 

subpart E of part 52 or under subpart H 

of part 53 of this chapter, the Director, 

Office of Nuclear Reactor Regulation, 

shall publish in the Federal Register a 

notice of receipt of the submittal, 

inviting comments from interested 

persons within 60 days of publication or 

other time as may be specified, for 

consideration by the NRC staff and 

ACRS in their review. 

* * * * * 

■ 13. In § 2.202, revise paragraph (e) to 

read as follows: 

§ 2.202 Orders. 

* * * * * 

(e)(1) If the order involves the 

modification of a part 50 or a part 53 

license and is a backfit, the 

requirements of § 50.109 or § 53.1590 of 

this chapter, as applicable, shall be 

followed, unless the licensee has 

consented to the action required. 

(2) If the order involves the 

modification of combined license under 

subpart C of part 52, or subpart H of part 

53 of this chapter, the requirements of 

§ 52.98 or § 53.1443 of this chapter, as 

applicable, shall be followed unless the 

licensee has consented to the action 

required. 

(3) If the order involves a change to 

an early site permit under subpart A of 

part 52 or under subpart H of part 53 of 

this chapter, the requirements of § 52.39 

or § 53.1188 of this chapter, as 

applicable, must be followed, unless the 

applicant or licensee has consented to 

the action required. 

(4) If the order involves a change to 

a standard design certification rule 

referenced by that plant’s application, 

the requirements, if any, in the 

referenced design certification rule with 

respect to changes must be followed, or, 

in the absence of these requirements, 


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the requirements of § 52.63 or § 53.1263 

of this chapter, as applicable, must be 

followed, unless the applicant or 

licensee has consented to follow the 

action required. 

(5) If the order involves a change to 

a standard design approval referenced 

by that plant’s application, the 

requirements of § 52.145 or § 53.1221 of 

this chapter, as applicable, must be 

followed unless the applicant or 

licensee has consented to follow the 

action required. 

(6) If the order involves a 

modification of a manufacturing license 

under subpart F of part 52 or under 

subpart H of part 53 of this chapter, the 

requirements of § 52.171 or § 53.1288 of 

this chapter, as applicable, must be 

followed, unless the applicant or 

licensee has consented to the action 

required. 

■ 14. In § 2.309, revise paragraphs (a), 

(f)(1)(i), (f)(1)(vi) and (vii), (g), (h)(2), 

(i)(2), (j) to read as follows: 

§ 2.309 Hearing requests, petitions to 

intervene, requirements for standing, and 

contentions. 

(a) General requirements. Any person 

whose interest may be affected by a 

proceeding and who desires to 

participate as a party must file a written 

request for hearing and a specification 

of the contentions which the person 

seeks to have litigated in the hearing. In 

a proceeding under § 52.103 or 

§ 53.1452 of this chapter, as applicable, 

the Commission, acting as the presiding 

officer, will grant the request if it 

determines that the requestor has 

standing under the provisions of 

paragraph (d) of this section and has 

proposed at least one admissible 

contention that meets the requirements 

of paragraph (f) of this section. For all 

other proceedings, except as provided in 

paragraph (e) of this section, the 

Commission, presiding officer, or the 

Atomic Safety and Licensing Board 

designated to rule on the request for 

hearing and/or petition for leave to 

intervene, will grant the request/petition 

if it determines that the requestor/ 

petitioner has standing under the 

provisions of paragraph (d) of this 

section and has proposed at least one 

admissible contention that meets the 

requirements of paragraph (f) of this 

section. In ruling on the request for 

hearing/petition to intervene submitted 

by petitioners seeking to intervene in 

the proceeding on the HLW repository, 

the Commission, the presiding officer, 

or the Atomic Safety and Licensing 

Board shall also consider any failure of 

the petitioner to participate as a 

potential party in the pre-license 

application phase under subpart J of this 

part in addition to the factors in 

paragraph (d) of this section. If a request 

for hearing or petition to intervene is 

filed in response to any notice of 

hearing or opportunity for hearing, the 

applicant/licensee shall be deemed to be 

a party. 

* * * * * 

(f) * * * 

(1) * * * 

(i) Provide a specific statement of the 

issue of law or fact to be raised or 

controverted, provided further, that the 

issue of law or fact to be raised in a 

request for hearing under § 52.103(b) or 

§ 53.1452(b) of this chapter, as 

applicable, must be directed at 

demonstrating that one or more of the 

acceptance criteria in the combined 

license have not been, or will not be 

met, and that the specific operational 

consequences of nonconformance 

would be contrary to providing 

reasonable assurance of adequate 

protection of the public health and 

safety; 

* * * * * 

(vi) In a proceeding other than one 

under § 52.103 or § 53.1452 of this 

chapter provide sufficient information 

to show that a genuine dispute exists 

with the applicant/licensee on a 

material issue of law or fact. This 

information must include references to 

specific portions of the application 

(including the applicant’s 

environmental report and safety report) 

that the petitioner disputes and the 

supporting reasons for each dispute, or, 

if the petitioner believes that the 

application fails to contain information 

on a relevant matter as required by law, 

the identification of each failure and the 

supporting reasons for the petitioner’s 

belief; and 

(vii) In a proceeding under § 52.103(b) 

or § 53.1452(b) of this chapter, as 

applicable, the information must be 

sufficient, and include supporting 

information showing, prima facie, that 

one or more of the acceptance criteria in 

the combined license have not been, or 

will not be met, and that the specific 

operational consequences of 

nonconformance would be contrary to 

providing reasonable assurance of 

adequate protection of the public health 

and safety. This information must 

include the specific portion of the report 

required by § 52.99(c) or § 53.1449(c) of 

this chapter, as applicable, which the 

requestor believes is inaccurate, 

incorrect, and/or incomplete (i.e., fails 

to contain the necessary information 

required by § 52.99(c) or § 53.1449(c) of 

this chapter, as applicable). If the 

requestor identifies a specific portion of 

the report under § 52.99(c) or 

§ 53.1449(c) of this chapter, as 

applicable, as incomplete and the 

requestor contends that the incomplete 

portion prevents the requestor from 

making the necessary prima facie 

showing, then the requestor must 

explain why this deficiency prevents 

the requestor from making the prima 

facie showing. 

* * * * * 

(g) Selection of hearing procedures. A 

request for hearing and/or petition for 

leave to intervene may, except in a 

proceeding under § 52.103 or § 53.1452 

of this chapter, as applicable, also 

address the selection of hearing 

procedures, taking into account the 

provisions of § 2.310. If a request/ 

petition relies upon § 2.310(d), the 

request/petition must demonstrate, by 

reference to the contention and the 

bases provided and the specific 

procedures in subpart G of this part, that 

resolution of the contention necessitates 

resolution of material issues of fact 

which may be best determined through 

the use of the identified procedures. 

(h) * * * 

(2) If the proceeding pertains to a 

production or utilization facility (as 

defined in § 50.2 or § 53.020 of this 

chapter) located within the boundaries 

of the State, local governmental body, or 

Federally-recognized Indian Tribe 

seeking to participate as a party, no 

further demonstration of standing is 

required. If the production or utilization 

facility is not located within the 

boundaries of the State, local 

governmental body, or Federallyrecognized Indian Tribe seeking to 

participate as a party, the State, local 

governmental body, or Federallyrecognized Indian Tribe also must 

demonstrate standing. 

* * * * * 

(i) * * * 

(2) Except in a proceeding under 

§ 52.103 or § 53.1452 of this chapter, as 

applicable, the participant who filed the 

hearing request, intervention petition, or 

motion for leave to file new or amended 

contentions after the deadline may file 

a reply to any answer. The reply must 

be filed within 7 days after service of 

that answer. 

* * * * * 

(j) Decision on request/petition. (1) In 

all proceedings other than a proceeding 

under § 52.103 or § 53.1452 of this 

chapter, as applicable, the presiding 

officer shall issue a decision on each 

request for hearing or petition to 

intervene within 45 days of the 

conclusion of the initial pre-hearing 

conference or, if no pre-hearing 

conference is conducted, within 45 days 

after the filing of answers and replies 


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under paragraph (i) of this section. With 

respect to a request to admit amended 

or new contentions, the presiding officer 

shall issue a decision on each such 

request within 45 days of the conclusion 

of any pre-hearing conference that may 

be conducted regarding the proposed 

amended or new contentions or, if no 

pre-hearing conference is conducted, 

within 45 days after the filing of 

answers and replies, if any. In the event 

the presiding officer cannot issue a 

decision within 45 days, the presiding 

officer shall issue a notice advising the 

Commission and the parties, and the 

notice shall include the expected date of 

when the decision will issue. 

(2) The Commission, acting as the 

presiding officer, shall expeditiously 

grant or deny the request for hearing in 

a proceeding under § 52.103 or 

§ 53.1452 of this chapter, as applicable. 

The Commission’s decision may not be 

the subject of any appeal under § 2.311. 

■ 15. Amend § 2.310 by: 

■ a. In paragraphs (a) and (h) 

introductory text, removing the crossreference ‘‘parts 30, 32 through 36, 39, 

40, 50, 52, 54, 55, 61, 70 and 72 of this 

chapter’’ and adding, in its place, the 

cross reference ‘‘parts 30, 32 through 36, 

39, 40, 50, 52, 53, 54, 55, 61, 70 and 72 

of this chapter’’; and 

■ b. Revising paragraphs (i) and (j). 

The revisions read as follows. 

§ 2.310 Selection of hearing procedures. 

* * * * * 

(i) In design certification rulemaking 

proceedings under part 52 or part 53 of 

this chapter, any informal hearing held 

under § 52.51 or § 53.1242 of this 

chapter, as applicable, must be 

conducted under the procedures of 

subpart O of this part. 

(j) Proceedings on a Commission 

finding under § 52.103(c) and (g) or 

§ 53.1452(c) and (g) of this chapter, as 

applicable, shall be conducted in 

accordance with the procedures 

designated by the Commission in each 

proceeding. 

* * * * * 

■ 16. In § 2.329, revise paragraph (a) to 

read as follows: 

§ 2.329 Prehearing conference. 

(a) Necessity for prehearing 

conference; timing. The Commission or 

the presiding officer may, and in the 

case of a proceeding on an application 

for a construction permit or an operating 

license for a facility of a type described 

in §§ 50.21(b) or 50.22, or part 53 of this 

chapter, or a testing facility, must direct 

the parties or their counsel to appear at 

a specified time and place for a 

conference or conferences before trial. A 

prehearing conference in a proceeding 

involving a construction permit or 

operating license for a facility of a type 

described in §§ 50.21(b) or 50.22 or part 

53 of this chapter must be held within 

sixty (60) days after discovery has been 

completed or any other time specified 

by the Commission or the presiding 

officer. 

* * * * * 

■ 17. In § 2.339, revise paragraph (d) to 

read as follows: 

§ 2.339 Expedited decision-making 

procedure. 

* * * * * 

(d) The provisions of this section do 

not apply to an initial decision directing 

the issuance of a limited work 

authorization under 10 CFR 50.10 or 

10 CFR 53.1130; an early site permit 

under subpart A of part 52 or under 

subpart H of part 53 of this chapter; a 

construction permit or construction 

authorization under part 50 or part 53 

of this chapter; a combined license 

under subpart C of part 52 or under 

subpart H of part 53 of this chapter; or 

a manufacturing license under subpart F 

of part 52 or under subpart H of part 53. 

■ 18. In § 2.340, revise paragraphs (b), 

(c), (d), (f), (i), and (j) to read as follows: 

§ 2.340 Initial decision in certain contested 

proceedings; immediate effectiveness of 

initial decisions; issuance of authorizations, 

permits and licenses. 

* * * * * 

(b) Initial decision—combined license 

under 10 CFR parts 52 or 53. (1) Matters 

in controversy; presiding officer 

consideration of matters not put in 

controversy by parties. In any initial 

decision in a contested proceeding on 

an application for a combined license 

under parts 52 or 53 of this chapter 

(including an amendment to or renewal 

of combined license), the presiding 

officer shall make findings of fact and 

conclusions of law on the matters put 

into controversy by the parties and any 

matter designated by the Commission to 

be decided by the presiding officer. The 

presiding officer shall also make 

findings of fact and conclusions of law 

on any matter not put into controversy 

by the parties, but only to the extent that 

the presiding officer determines that a 

serious safety, environmental, or 

common defense and security matter 

exists, and the Commission approves of 

an examination of and decision on the 

matter upon its referral by the presiding 

officer under, inter alia, the provisions 

of §§ 2.323 and 2.341. 

(2) Presiding officer initial decision 

and issuance of permit or license. 

(i) In a contested proceeding for the 

initial issuance or renewal of a 

combined license under parts 52 or 53 

of this chapter, or the amendment of a 

combined license where the NRC has 

not made a determination of no 

significant hazards consideration, the 

Commission or the Director, Office of 

Nuclear Reactor Regulation, as 

appropriate after making the requisite 

findings, shall issue, deny, or 

appropriately condition the permit or 

license in accordance with the presiding 

officer’s initial decision once that 

decision becomes effective. 

(ii) In a contested proceeding for the 

amendment of a combined license 

under parts 52 or 53 of this chapter 

where the NRC has made a 

determination of no significant hazards 

consideration, the Commission or the 

Director, Office of Nuclear Reactor 

Regulation, as appropriate (appropriate 

official), after making the requisite 

findings and complying with any 

applicable provisions of § 2.1202(a) or 

§ 2.1403(a), may issue the amendment 

before the presiding officer’s initial 

decision becomes effective. Once the 

presiding officer’s initial decision 

becomes effective, the appropriate 

official shall take action with respect to 

that amendment in accordance with the 

initial decision. If the presiding officer’s 

initial decision becomes effective before 

the appropriate official issues the 

amendment, then the appropriate 

official, after making the requisite 

findings, shall issue, deny, or 

appropriately condition the amendment 

in accordance with the presiding 

officer’s initial decision. 

(c) Initial decision on findings under 

10 CFR 52.103 or 10 CFR 53.1452 with 

respect to acceptance criteria in nuclear 

power reactor combined licenses. In any 

initial decision under § 52.103(g) or 

§ 53.1452(g) of this chapter with respect 

to whether acceptance criteria have 

been or will be met, the presiding officer 

shall make findings of fact and 

conclusions of law on the matters put 

into controversy by the parties, and any 

matter designated by the Commission to 

be decided by the presiding officer. 

Matters not put into controversy by the 

parties but identified by the presiding 

officer as matters requiring further 

examination, shall be referred to the 

Commission for its determination; the 

Commission may, in its discretion, treat 

any of these referred matters as a request 

for action under § 2.206 and process the 

matter in accordance with § 52.103(f) or 

§ 53.1452(f) of this chapter. 

(d) Initial decision—manufacturing 

license under 10 CFR parts 52 or 53. (1) 

Matters in controversy; presiding officer 

consideration of matters not put in 

controversy by parties. In any initial 

decision in a contested proceeding on 

an application for a manufacturing 


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license under subpart C of part 52 or 

subpart H of part 53 of this chapter 

(including an amendment to or renewal 

of a manufacturing license), the 

presiding officer shall make findings of 

fact and conclusions of law on the 

matters put into controversy by the 

parties and any matter designated by the 

Commission to be decided by the 

presiding officer. The presiding officer 

also shall make findings of fact and 

conclusions of law on any matter not 

put into controversy by the parties, but 

only to the extent that the presiding 

officer determines that a serious safety, 

environmental, or common defense and 

security matter exists, and the 

Commission approves of an 

examination of and decision on the 

matter upon its referral by the presiding 

officer under, inter alia, the provisions 

of §§ 2.323 and 2.341. 

(2) Presiding officer initial decision 

and issuance of permit or license. 

(i) In a contested proceeding for the 

initial issuance or renewal of a 

manufacturing license under subpart C 

of part 52 or subpart H of part 53 of this 

chapter, or the amendment of a 

manufacturing license, the Commission 

or the Director, Office of Nuclear 

Reactor Regulation, as appropriate, after 

making the requisite findings, shall 

issue, deny, or appropriately condition 

the permit or license in accordance with 

the presiding officer’s initial decision 

once that decision becomes effective. 

(ii) In a contested proceeding for the 

initial issuance or renewal of a 

manufacturing license under subpart C 

of part 52 or subpart H of part 53 of this 

chapter, or the amendment of a 

manufacturing license, the Commission 

or the Director, Office of Nuclear 

Reactor Regulation, as appropriate 

(appropriate official), may issue the 

license, permit, or license amendment 

in accordance with § 2.1202(a) or 

§ 2.1403(a) before the presiding officer’s 

initial decision becomes effective. If, 

however, the presiding officer’s initial 

decision becomes effective before the 

license, permit, or license amendment is 

issued under § 2.1202 or § 2.1403, then 

the Commission or the Director, Office 

of Nuclear Reactor Regulation, as 

appropriate, shall issue, deny, or 

appropriately condition the license, 

permit, or license amendment in 

accordance with the presiding officer’s 

initial decision. 

* * * * * 

(f) Immediate effectiveness of certain 

presiding officer decisions. A presiding 

officer’s initial decision directing the 

issuance or amendment of a limited 

work authorization under § 50.10 or 

§ 53.1130 of this chapter; an early site 

permit under subpart A of part 52 or 

under subpart H of part 53 of this 

chapter; a construction permit or 

construction authorization under part 

50 or part 53 of this chapter; an 

operating license under part 50 or part 

53 of this chapter; a combined license 

under subpart C of part 52 or subpart H 

or part 53 of this chapter; a 

manufacturing license under subpart F 

of part 52 or subpart H of part 53 of this 

chapter; a renewed license under part 

53 or part 54 of this chapter; or a license 

under part 72 of this chapter to store 

spent fuel in an independent spent fuel 

storage facility (ISFSI) or a monitored 

retrievable storage installation (MRS); 

an initial decision directing issuance of 

a license under part 61 of this chapter; 

or an initial decision under § 52.103(g) 

or § 53.1452(g) of this chapter that 

acceptance criteria in a combined 

license have been met, is immediately 

effective upon issuance unless the 

presiding officer finds that good cause 

has been shown by a party why the 

initial decision should not become 

immediately effective. 

* * * * * 

(i) Issuance of authorizations, 

permits, and licenses—production and 

utilization facilities. The Commission or 

the Director, Office of Nuclear Reactor 

Regulation, as appropriate, shall issue a 

limited work authorization under 

§ 50.10 or § 53.1130 of this chapter; an 

early site permit under subpart A of part 

52 or subpart H of part 53 of this 

chapter; a construction permit or 

construction authorization under part 

50 or part 53 of this chapter; an 

operating license under part 50 or part 

53 of this chapter; a combined license 

under subpart C of part 52 or part 53 of 

this chapter; or a manufacturing license 

under subpart F of part 52 or part 53 of 

this chapter within 10 days from the 

date of issuance of the initial decision: 

(1) If the Commission or the Director 

has made all findings necessary for 

issuance of the authorization, permit or 

license, not within the scope of the 

initial decision of the presiding officer; 

and 

(2) Notwithstanding the pendency of 

a petition for reconsideration under 

§ 2.345, a petition for review under 

§ 2.341, or a motion for stay under 

§ 2.342, or the filing of a petition under 

§ 2.206. 

(j) Issuance of finding on acceptance 

criteria under 10 CFR 52.103 or 10 CFR 

53.1452. The Commission or the 

Director, Office of Nuclear Reactor 

Regulation, as appropriate, shall make 

the finding under § 52.103(g) or 

§ 53.1452(g) of this chapter, that 

acceptance criteria in a combined 

license are met within 10 days from the 

date of the presiding officer’s initial 

decision: 

(1) If the Commission or the Director 

is otherwise able to make the finding 

under § 52.103(g) or § 53.1452(g) of this 

chapter, that the prescribed acceptance 

criteria are met for those acceptance 

criteria not within the scope of the 

initial decision of the presiding officer; 

(2) If the presiding officer’s initial 

decision—with respect to contentions 

that the prescribed acceptance criteria 

have not been met—finds that those 

acceptance criteria have been met, and 

the Commission or the Director 

thereafter is able to make the finding 

that those acceptance criteria are met; 

(3) If the presiding officer’s initial 

decision—with respect to contentions 

that the prescribed acceptance criteria 

will not be met—finds that those 

acceptance criteria will be met, and the 

Commission or the Director thereafter is 

able to make the finding that those 

acceptance criteria are met; and 

(4) Notwithstanding the pendency of 

a petition for reconsideration under 

§ 2.345, a petition for review under 

§ 2.341, or a motion for stay under 

§ 2.342, or the filing of a petition under 

§ 2.206. 

* * * * * 

§ 2.341 [Amended] 

■ 19. In § 2.341(a)(1), remove the phrase 

‘‘§ 52.103(c)’’ and add in its place the 

phrase ‘‘§ 52.103(c) or § 53.1452(c)’’. 

§ 2.400 [Amended] 

■ 20. In § 2.400, remove the phrase 

‘‘parts 50 or 52’’ and add in its place the 

phrase ‘‘part 50 or part 52, or 

§ 53.1470’’. 

■ 21. In § 2.401, revise the section 

heading and paragraph (a) to read as 

follows: 

§ 2.401 Notice of hearing on construction 

permit or combined license applications 

pursuant to appendix N of 10 CFR parts 50, 

52, or 53. 

(a) In the case of applications under 

appendix N of part 50 or § 53.1470 of 

this chapter for construction permits for 

nuclear power reactors of the type 

described in § 50.22 or part 53 of this 

chapter, or applications under appendix 

N of part 52 or § 53.1470 of this chapter 

for combined licenses, the Secretary 

will issue notices of hearing pursuant to 

§ 2.104. 

* * * * * 

■ 22. In § 2.402, revise paragraph (a) to 

read as follows: 

§ 2.402 Separate hearings on separate 

issues; consolidation of proceedings. 

(a) In the case of applications under 

appendix N of part 50 or § 53.1470 of 


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this chapter for construction permits for 

nuclear power reactors of a type 

described in 10 CFR 50.22 or part 53, or 

applications pursuant to appendix N of 

part 52 or § 53.1470 of this chapter for 

combined licenses, the Commission or 

the presiding officer may order separate 

hearings on particular phases of the 

proceeding, such as matters related to 

the acceptability of the design of the 

reactor in the context of the site 

parameters postulated for the design or 

environmental matters. 

* * * * * 

§ 2.403 [Amended] 

■ 23. In § 2.403, remove the phrase 

‘‘appendix N of part 50’’ and add in its 

place the phrase ‘‘appendix N to part 50 

or § 53.1470’’. 

§ 2.404 [Amended] 

■ 24. In § 2.404, remove the phrase 

‘‘appendix N of part 50’’ and add in its 

place the phrase ‘‘appendix N to part 50 

or § 53.1470’’. 

§ 2.405 [Amended] 

■ 25. In § 2.405, remove the phrase 

‘‘part 52’’ and add in its place the 

phrase ‘‘part 52 or part 53’’. 

§ 2.406 [Amended] 

■ 26. In § 2.406, remove the phrase 

‘‘appendix N of parts 50 or 52’’ and add 

in its place the phrase ‘‘appendix N to 

part 50 or part 52 or § 53.1470’’. 

§ 2.500 [Amended] 

■ 27. In § 2.500, remove the phrase 

‘‘subpart F of part 52’’ and add in its 

place the phrase ‘‘subpart F of part 52 

or subpart H of part 53’’. 

■ 28. In § 2.501, revise the section 

heading and paragraph (a) introductory 

text to read as follows: 

§ 2.501 Notice of hearing on application 

under 10 CFR parts 52 or 53 for a license 

to manufacture nuclear power reactors. 

(a) In the case of an application under 

subpart F of part 52 or subpart H of part 

53 of this chapter for a license to 

manufacture nuclear power reactors of 

the type described in § 50.22 or part 53 

of this chapter to be operated at sites not 

identified in the license application, the 

Secretary will issue a notice of hearing 

to be published in the Federal Register 

at least 30 days before the date set for 

hearing in the notice.1 The notice shall 

be issued as soon as practicable after the 

application has been docketed. The 

notice will state: 

* * * * * 1The 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. 

■ 29. In § 2.643, revise paragraph (b) to 

read as follows: 

§ 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 § 51.20(b) of 

this chapter and is of the type specified 

in § 50.21(b)(2) or (3) or § 50.22 or part 

53 of this chapter or an application for 

a combined license where part one of 

the application as described in 

§ 2.101(a)(9) is complete. Part one will 

not be considered complete unless it 

contains the information required by 

§ 50.10(d)(3) or § 53.1130(a)(3) of this 

chapter. Upon assignment of a docket 

number, the procedures in § 2.101(a)(3) 

and (4) relating to formal docketing and 

the submission and distribution of 

additional copies of the application 

must be followed. 

* * * * * 

§ 2.645 [Amended] 

■ 30. In § 2.645, in paragraph (a), 

remove the phrase ‘‘§ 50.33(a) through 

(f) of this chapter’’ and add in its place 

the phrase ‘‘§§ 50.33(a) through (f), 

53.1109, and 53.1306(a) or 53.1413 of 

this chapter, as applicable,’’. 

§ 2.649 [Amended] 

■ 31. In § 2.649, remove the phrase ‘‘10 

CFR 50.10(d)’’ and add in its place the 

phrase ‘‘10 CFR 50.10(d) or 10 CFR 

53.1130(a)’’. 

§ 2.800 [Amended] 

■ 32. In § 2.800, amend paragraphs (c) 

and (d) by removing the phrase ‘‘subpart 

B of part 52’’ and adding in its place the 

phrase ‘‘subpart B of part 52 or subpart 

H of part 53’’. 

§ 2.801 [Amended] 

■ 33. In § 2.801, remove the phrase 

‘‘subpart B of part 52’’ and add in its 

place the phrase ‘‘subpart B of part 52 

or subpart H of part 53’’. 

§ 2.813 [Amended] 

■ 34. In § 2.813(a), remove the phrase 

‘‘parts 50, 52, and 100’’ and add in its 

place the phrase ‘‘parts 50, 52, 53, and 

100’’. 

§ 2.1103 [Amended] 

■ 35. In § 2.1103, remove the phrase 

‘‘part 50 of this chapter’’ and add in its 

place the phrase ‘‘parts 50 or 53 of this 

chapter’’. 

■ 36. In § 2.1202, revise paragraphs 

(a)(1) through (3) and (a)(6) to read as 

follows: 

§ 2.1202 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 

§§ 50.12 or 53.1130 of this chapter, or an 

application for a combined license 

under subpart C of 10 CFR part 52, or 

under subpart H of 10 CFR part 53; 

(2) An application for an early site 

permit under subpart A of 10 CFR part 

52 or under subpart H of 10 CFR part 

53; 

(3) An application for a 

manufacturing license under subpart F 

of 10 CFR part 52 or under subpart H 

of 10 CFR part 53; 

* * * * * 

(6) Production or utilization facility 

licensing actions that involve significant 

hazards considerations as defined in 

§§ 50.92 or 53.1520 of this chapter. 

* * * * * 

§ 2.1301 [Amended] 

■ 37. In § 2.1301(b), remove ‘‘part 50 

and part 52’’ and add in its place ‘‘parts 

50, 52, and 53’’. 

§ 2.1403 [Amended] 

■ 38. In § 2.1403, remove the phrase ‘‘10 

CFR 50.92’’ and add in its place the 

phrase ‘‘10 CFR 50.92 or 10 CFR 

53.1520’’. 

§ 2.1500 [Amended] 

■ 39. In § 2.1500, remove the phrase 

‘‘subpart B of part 52’’ and add in its 

place the phrase ‘‘subpart B of part 52 

or under subpart H of part 53’’. 

§ 2.1502 [Amended] 

■ 40. In § 2.1502, in paragraph (a), 

remove the phrase ‘‘§ 52.51(b)’’ and add 

in its place the phrase ‘‘§§ 52.51(b) or 

53.1242(b)(2)’’; and in paragraph (b)(1), 

wherever it appears, remove the phrase 

‘‘§ 52.51(a)’’ and add in its place the 

phrase ‘‘§§ 52.51(a) or 53.1242(b)’’. 

PART 10—CRITERIA AND 

PROCEDURES FOR DETERMINING 

ELIGIBILITY FOR ACCESS TO 

RESTRICTED DATA OR NATIONAL 

SECURITY INFORMATION OR AN 

EMPLOYMENT CLEARANCE 

■ 41. The authority citation for part 10 

continues to read as follows: 

Authority: Atomic Energy Act of 1954, 

secs. 145, 161 (42 U.S.C. 2165, 2201); Energy 

Reorganization Act of 1974, sec. 201 (42 

U.S.C. 5841); E.O. 10450, 18 FR 2489, 3 CFR, 

1949–1953 Comp., p. 936, as amended; E.O. 

10865, 25 FR 1583, 3 CFR, 1959–1963 Comp., 

p. 398, as amended; E.O. 12968, 60 FR 40245, 

3 CFR, 1995 Comp., p. 391. 

§ 10.1 [Amended] 

■ 42. In § 10.1, in paragraph (a)(3) 

remove the phrase ‘‘under part 52’’ and 


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add in its place the phrase ‘‘under parts 

52 or 53’’. 

§ 10.2 [Amended] 

■ 43. In § 10.2, in paragraph (b), 

wherever it appears, remove the phrase 

‘‘under part 52’’ and add in its place the 

phrase ‘‘under parts 52 or 53’’. 

PART 11—CRITERIA AND 

PROCEDURES FOR DETERMINING 

ELIGIBILITY FOR ACCESS TO OR 

CONTROL OVER SPECIAL NUCLEAR 

MATERIAL 

■ 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. 

§ 11.7 [Amended] 

■ 45. In § 11.7, in the introductory text, 

remove the phrase ‘‘parts 10, 25, 50, 70, 

72, 73, and 95 of this chapter’’ and add 

in its place the phrase ‘‘parts 10, 25, 50, 

53, 70, 72, 73, and 95 of this chapter’’. 

PART 19—NOTICES, INSTRUCTIONS 

AND REPORTS TO WORKERS: 

INSPECTION AND INVESTIGATIONS 

■ 46. The authority citation for part 19 

continues to read as follows: 

Authority: Atomic Energy Act of 1954, 

secs. 53, 63, 81, 103, 104, 161, 223, 234, 1701 

(42 U.S.C. 2073, 2093, 2111, 2133, 2134, 

2201, 2273, 2282, 2297f); Energy 

Reorganization Act of 1974, secs. 201, 211, 

401 (42 U.S.C. 5841, 5851, 5891); 44 U.S.C. 

3504 note. 

■ 47. In § 19.2, revise paragraph (a) to 

read as follows: 

§ 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. 

* * * * * 

■ 48. In § 19.3, revise the definitions for 

‘‘License’’ and ‘‘Regulated entities’’ to 

read as follows: 

§ 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. 

* * * * * 

§ 19.11 [Amended] 

■ 49. In § 19.11, in paragraph (a) 

introductory text, paragraph (b) 

introductory text, and paragraph (e)(1), 

remove the phrase ‘‘of part 52’’ 

wherever may appears and add in its 

place the phrase ‘‘of part 52 or under 

subpart H of part 53’’. 

§ 19.14 [Amended] 

■ 50. In § 19.14, in paragraph (a), 

wherever it may appear, remove the 

phrase ‘‘of part 52’’ and add in its place 

the phrase ‘‘of part 52 or under subpart 

H of part 53’’. 

§ 19.20 [Amended] 

■ 51. In § 19.20, add the number ‘‘53,’’ 

in sequential order. 

PART 20—STANDARDS FOR 

PROTECTION AGAINST RADIATION 

■ 52. The authority citation for part 20 

continues to read as follows: 

Authority: Atomic Energy Act of 1954, 

secs. 11, 53, 63, 65, 81, 103, 104, 161, 170H, 

182, 186, 223, 234, 274, 1701 (42 U.S.C. 2014, 

2073, 2093, 2095, 2111, 2133, 2134, 2201, 

2210h, 2232, 2236, 2273, 2282, 2021, 2297f); 

Energy Reorganization Act of 1974, secs. 201, 

202 (42 U.S.C. 5841, 5842); Low-Level 

Radioactive Waste Policy Amendments Act 

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

3504 note. 

§ 20.1002 [Amended] 

■ 53. In § 20.1002, remove the phrase 

‘‘parts 30 through 36, 39, 40, 50, 52, 60, 

61, 63, 70, or 72 of this chapter’’ and 

add in its place the phrase ‘‘parts 30 

through 36, 39, 40, 50, 52, 53, 60, 61, 

63, 70, or 72 of this chapter’’. 

■ 54. In § 20.1003, revise the definition 

for ‘‘License’’ to read as follows: 

§ 20.1003 Definitions. 

* * * * * 

License means a license issued under 

the regulations in parts 30 through 36, 

39, 40, 50, 53, 60, 61, 63, 70, or 72 of 

this chapter. 

* * * * * 

§ 20.1101 [Amended] 

■ 55. In § 20.1101, in paragraph (d), 

remove the phrase ‘‘subject to § 50.34a’’ 

and add in its place the phrase ‘‘subject 

to §§ 50.34a or 53.260 of this chapter’’. 

§ 20.1401 [Amended] 

■ 56. Amend § 20.1401 by: 

■ a. In paragraph (a), removing the 

phrase ‘‘parts 30, 40, 50, 52, 60, 61, 63, 

70, and 72 of this chapter’’, and adding 

in its place the phrase ‘‘parts 30, 40, 50, 

52, 53, 60, 61, 63, 70, and 72 of this 

chapter’’; and 

■ b. In paragraphs (a) and (c) removing 

the phrase ‘‘in accordance with § 50.83’’ 

and adding in its place the phrase ‘‘in 

accordance with §§ 50.83 or 53.1080’’. 

■ 57. In § 20.1403, revise paragraph (d) 

introductory text to read as follows: 

§ 20.1403 Criteria for license termination 

under restricted conditions. 

* * * * * 

(d) The licensee has submitted a 

decommissioning plan or License 

Termination Plan (LTP) to the 

Commission indicating the licensee’s 

intent to decommission in accordance 

with §§ 30.36(d), 40.42(d), 50.82 (a) and 

(b), subpart G of part 53, 70.38(d), or 

72.54 of this chapter, and specifying 

that the licensee intends to 

decommission by restricting use of the 

site. The licensee shall document in the 

LTP or decommissioning plan how the 

advice of individuals and institutions in 

the community who may be affected by 

the decommissioning has been sought 

and incorporated, as appropriate, 

following analysis of that advice. 

* * * * * 

■ 58. In § 20.1404, revise paragraph 

(a)(4) introductory text to read as 

follows: 

§ 20.1404 Alternate criteria for license 

termination. 

(a) * * * 


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(4) Has submitted a decommissioning 

plan or License Termination Plan (LTP) 

to the Commission indicating the 

licensee’s intent to decommission in 

accordance with § 30.36(d), 40.42(d), 

50.82 (a) and (b), subpart G of part 53, 

70.38(d), or 72.54 of this chapter, and 

specifying that the licensee proposes to 

decommission by use of alternate 

criteria. The licensee shall document in 

the decommissioning plan or LTP how 

the advice of individuals and 

institutions in the community who may 

be affected by the decommissioning has 

been sought and addressed, as 

appropriate, following analysis of that 

advice. In seeking such advice, the 

licensee shall provide for: 

* * * * * 

§ 20.1406 [Amended] 

■ 59. In § 20.1406, in paragraphs (a) and 

(b), wherever it appears, remove the 

phrase ‘‘under part 52’’ and add in its 

place the phrase ‘‘under parts 52 or 53’’. 

■ 60. In § 20.1501, revise paragraph (b) 

to read as follows: 

§ 20.1501 General. 

* * * * * 

(b) Notwithstanding § 20.2103(a) of 

this part, records from surveys 

describing the location and amount of 

subsurface residual radioactivity 

identified at the site must be kept with 

records important for decommissioning, 

and such records must be retained in 

accordance with § 30.35(g), § 40.36(f), 

§ 50.75(g), subpart G of part 53, 

§ 70.25(g), or § 72.30(d) of this chapter, 

as applicable. 

* * * * * 

§ 20.1905 [Amended] 

■ 61. In § 20.1905, in paragraph (g) 

introductory text, remove the phrase 

‘‘Parts 50 or 52’’ and add in its place the 

phrase ‘‘parts 50, 52, or 53’’. 

■ 62. In § 20.2004, revise paragraph 

(b)(1) to read as follows: 

§ 20.2004 Treatment or disposal by 

incineration. 

* * * * * 

(b)(1) Waste oils (petroleum derived 

or synthetic oils used principally as 

lubricants, coolants, hydraulic or 

insulating fluids, or metalworking oils) 

that have been radioactively 

contaminated in the course of the 

operation or maintenance of a nuclear 

power reactor licensed under parts 50 or 

53 of this chapter may be incinerated on 

the site where generated provided that 

the total radioactive effluents from the 

facility, including the effluents from 

such incineration, conform to the 

requirements of appendix I to part 50 or 

§ 53.425(d) of this chapter and the 

effluent release limits contained in 

applicable license conditions other than 

effluent limits specifically related to 

incineration of waste oil. The licensee 

shall report any changes or additions to 

the information supplied under 

§§ 50.34, 50.34a, or under subpart H of 

part 53 of this chapter associated with 

this incineration pursuant to §§ 50.71 or 

53.1620 of this chapter, as appropriate. 

The licensee shall also follow the 

procedures of §§ 50.59 or 53.1565 of this 

chapter with respect to such changes to 

the facility or procedures. 

* * * * * 

■ 63. In § 20.2201, revise paragraphs 

(a)(2)(i), (b)(2)(i), and (c) to read as 

follows: 

§ 20.2201 Reports of theft or loss of 

licensed material. 

(a) * * * 

(2) * * * 

(i) Licensees having an installed 

Emergency Notification System shall 

make the reports to the NRC Operations 

Center under §§ 50.72 or 53.1630 of this 

chapter, and 

* * * * * 

(b) * * * 

(2) * * * 

(i) For holders of an operating license 

for a nuclear power plant, the events 

included in paragraph (b) of this section 

must be reported under the procedures 

described in §§ 50.73(b), (c), (d), (e), and 

(g) or 53.1640(b), (c), (d), and (e) of this 

chapter and must include the 

information required in paragraph (b)(1) 

of this section, and 

* * * * * 

(c) A duplicate report is not required 

under paragraph (b) of this section if the 

licensee is also required to submit a 

report pursuant to §§ 30.55(c), 37.57, 

37.81, 40.64(c), 50.72, 50.73, 53.1630, 

53.1640, 70.52, 73.27(b), 73.67(e)(3)(vii), 

73.67(g)(3)(iii), 73.1205, or 150.19(c) of 

this chapter. 

* * * * * 

§ 20.2202 [Amended] 

■ 64. In § 20.2202, in paragraph (d)(1), 

remove the phrase ‘‘10 CFR 50.72’’ and 

add in its place the phrase ‘‘§§ 50.72 or 

53.1630 of this chapter;’’. 

■ 65. In § 20.2203, revise paragraph (c) 

to read as follows: 

§ 20.2203 Reports of exposures, radiation 

levels, and concentrations of radioactive 

material exceeding the constraints or limits. 

* * * * * 

(c) For holders of an operating license 

or a combined license for a nuclear 

power plant, the occurrences included 

in paragraph (a) of this section must be 

reported under the procedures 

described in §§ 50.73(b), (c), (d), (e), and 

(g) or 53.1640(b), (c), (d), and (e) of this 

chapter, and must include the 

information required by paragraph (b) of 

this section. Occurrences reported 

under §§ 50.73 or 53.1640 of this 

chapter need not be reported by a 

duplicate report under paragraph (a) of 

this section. 

* * * * * 

§ 20.2206 [Amended] 

■ 66. In § 20.2206, in paragraph (a)(1), 

remove the phrase ‘‘or § 50.22’’ and add 

in its place the phrase ‘‘, § 50.22, or part 

53’’. 

PART 21—REPORTING OF DEFECTS 

AND NONCOMPLIANCE 

■ 67. The authority citation for part 21 

continues to read as follows: 

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. 

■ 68. In § 21.2, revise paragraphs (a)(2) 

through (4), (b), and (c) to read as 

follows: 

§ 21.2 Scope. 

(a) * * * 

(1) * * * 

(2) Each individual, corporation, 

partnership, or other entity doing 

business within the United States, and 

each director and responsible officer of 

such an organization, that constructs a 

production or utilization facility 

licensed for manufacture, construction, 

or operation under parts 50, 52, or 53 of 

this chapter, an ISFSI for the storage of 

spent fuel licensed under part 72 of this 

chapter, an MRS for the storage of spent 

fuel or high-level radioactive waste 

under part 72 of this chapter, or a 

geologic repository for the disposal of 

high-level radioactive waste under parts 

60 or 63 of this chapter; or supplies 

basic components for a facility or 

activity licensed, other than for export, 

under parts 30, 40, 50, 52, 53, 60, 61, 63, 

70, 71, or 72 of this chapter; 

(3) Each individual, corporation, 

partnership, or other entity doing 

business within the United States, and 

each director and responsible officer of 

such an organization, applying for a 

design certification rule under parts 52 

or 53 of this chapter; or supplying basic 

components with respect to that design 

certification, and each individual, 

corporation, partnership, or other entity 

doing business within the United States, 

and each director and responsible 

officer of such an organization, whose 


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application for design certification has 

been granted under parts 52 or 53 of this 

chapter, or who has supplied or is 

supplying basic components with 

respect to that design certification; 

(4) Each individual, corporation, 

partnership, or other entity doing 

business within the United States, and 

each director and responsible officer of 

such an organization, applying for or 

holding a standard design approval 

under parts 52 or 53 of this chapter; or 

supplying basic components with 

respect to a standard design approval 

under parts 52 or 53 of this chapter; 

(b) For persons licensed to construct 

a facility under either a construction 

permit issued under §§ 50.23 or 53.1333 

of this chapter or a combined license 

under parts 52 or 53 of this chapter (for 

the period of construction until the date 

that the Commission makes the finding 

under §§ 52.103(g) or 53.1452(g) of this 

chapter), or to manufacture a facility 

under parts 52 or 53 of this chapter, 

evaluation of potential defects and 

failures to comply and reporting of 

defects and failures to comply under 

§§ 50.55(e) or 53.605 of this chapter 

satisfies each person’s evaluation, 

notification, and reporting obligation to 

report defects and failures to comply 

under this part and the responsibility of 

individual directors and responsible 

officers of these licensees to report 

defects under Section 206 of the Energy 

Reorganization Act of 1974. 

(c) For persons licensed to operate a 

nuclear power plant under part 50, part 

52, or part 53 of this chapter, evaluation 

of potential defects and appropriate 

reporting of defects under §§ 50.72, 

50.73, 53.1630, 53.1640, or 73.1200 and 

73.1205 of this chapter, satisfies each 

person’s evaluation, notification, and 

reporting obligation to report defects 

under this part, and the responsibility of 

individual directors and responsible 

officers of these licensees to report 

defects under Section 206 of the Energy 

Reorganization Act of 1974. 

* * * * * 

■ 69. In § 21.3, revise the definitions for 

‘‘Basic component’’, ‘‘Commercial grade 

item’’, ‘‘Critical characteristics’’, 

‘‘Dedicating entity’’, ‘‘Dedication’’, 

‘‘Defect’’, and ‘‘Substantial safety 

hazard’’ to read as follows: 

§ 21.3 Definitions. 

* * * * * 

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 safeshutdown condition; or 

(C) The capability to prevent or 

mitigate the consequences of accidents 

which could result in potential offsite 

exposures comparable to those referred 

to in §§ 50.34(a)(1), 50.67(b)(2), or 

100.11 of this chapter, as applicable. 

(ii) Basic components are items 

designed and manufactured under a 

quality assurance program complying 

with appendix B to part 50 of this 

chapter, or commercial grade items 

which have successfully completed the 

dedication process. 

(2) When applied to standard design 

certifications and approvals under part 

53 of this chapter, basic component 

means the design or procurement 

information approved or to be approved 

within the scope of the design 

certification or approval for a safetyrelated SSC, or part thereof. When 

applied to standard design certifications 

under subpart B of part 52 of this 

chapter and standard design approvals 

under part 52 of this chapter, basic 

component means the design or 

procurement information approved or to 

be approved within the scope of the 

design certification or approval for an 

SSC, or part thereof, that affects its 

safety function necessary to assure: 

(i) The integrity of the reactor coolant 

pressure boundary; 

(ii) The capability to shut down the 

reactor and maintain it in a safeshutdown condition; or 

(iii) The capability to prevent or 

mitigate the consequences of accidents 

which could result in potential offsite 

exposures comparable to those referred 

to in §§ 50.34(a)(1), 50.67(b)(2), or 

100.11 of this chapter, as applicable. 

(3) When applied to other facilities 

and other activities licensed under 10 

CFR parts 30, 40, 50 (other than nuclear 

power plants), 60, 61, 63, 70, 71, or 72 

of this chapter, basic component means 

a structure, system, or component, or 

part thereof, that affects their safety 

function, that is directly procured by the 

licensee of a facility or activity subject 

to the regulations in this part and in 

which a defect or failure to comply with 

any applicable regulation in this 

chapter, order, or license issued by the 

Commission could create a substantial 

safety hazard. 

(4) In all cases, basic component 

includes safety-related design, analysis, 

inspection, testing, fabrication, 

replacement of parts, or consulting 

services that are associated with the 

component hardware, design 

certification, design approval, or 

information in support of an early site 

permit application under part 52 or part 

53 of this chapter, whether these 

services are performed by the 

component supplier or others. 

Commercial grade item. (1) When 

applied to nuclear power plants 

licensed under parts 50 or 53 of this 

chapter, commercial grade item means 

an SSC, or part thereof that affects its 

safety function, that was not designed 

and manufactured as a basic 

component. Commercial grade items do 

not include items where the design and 

manufacturing process require inprocess inspections and verifications to 

ensure that defects or failures to comply 

are identified and corrected (i.e., one or 

more critical characteristics of the item 

cannot be verified). 

(2) When applied to facilities and 

activities licensed pursuant to parts 30, 

40, 50 (other than nuclear power 

plants), 60, 61, 63, 70, 71, or 72 of this 

chapter, commercial grade item means 

an item that is: 

(i) Not subject to design or 

specification requirements that are 

unique to those facilities or activities; 

(ii) Used in applications other than 

those facilities or activities; and 

(iii) To be ordered from the 

manufacturer/supplier on the basis of 

specifications set forth in the 

manufacturer’s published product 

description (for example, a catalog). 

* * * * * 

Critical characteristics. When applied 

to nuclear power plants licensed under 

parts 50, 52, or 53 of this chapter, 

critical characteristics are those 

important design, material, and 

performance characteristics of a 

commercial grade item that, once 

verified, will provide reasonable 

assurance that the item will perform its 

intended safety function. 

Dedicating entity. When applied to 

nuclear power plants licensed under 

parts 50, 52, or 53 of this chapter, 

dedicating entity means the 

organization that performs the 

dedication process. Dedication may be 

performed by the manufacturer of the 

item, a third-party dedicating entity, or 

the licensee itself. The dedicating entity, 

under § 21.21(c) of this part, is 

responsible for identifying and 

evaluating deviations, reporting defects 

and failures to comply for the dedicated 

item, and maintaining auditable records 

of the dedication process. 

Dedication. (1) When applied to 

nuclear power plants licensed pursuant 

to 10 CFR parts 30, 40, 50, 53, or 60, 

dedication is an acceptance process 


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undertaken to provide reasonable 

assurance that a commercial grade item 

to be used as a basic component will 

perform its intended safety function 

and, in this respect, is deemed 

equivalent to an item designed and 

manufactured under a 10 CFR part 50, 

appendix B, quality assurance program. 

This assurance is achieved by 

identifying the critical characteristics of 

the item and verifying their 

acceptability by inspections, tests, or 

analyses performed by the purchaser or 

third-party dedicating entity after 

delivery, supplemented as necessary by 

one or more of the following: 

commercial grade surveys; product 

inspections or witness at holdpoints at 

the manufacturer’s facility, and analysis 

of historical records for acceptable 

performance. In all cases, the dedication 

process must be conducted under the 

applicable provisions of 10 CFR part 50, 

appendix B. The process is considered 

complete when the item is designated 

for use as a basic component. 

(2) When applied to facilities and 

activities licensed pursuant to 10 CFR 

parts 30, 40, 50 (other than nuclear 

power plants), 60, 61, 63, 70, 71, or 72, 

dedication occurs after receipt when 

that item is designated for use as a basic 

component. 

Defect means: 

(1) A deviation in a basic component 

delivered to a purchaser for use in a 

facility or an activity subject to the 

regulations in this part if, on the basis 

of an evaluation, the deviation could 

create a substantial safety hazard; 

(2) The installation, use, or operation 

of a basic component containing a 

defect as defined in this section; 

(3) A deviation in a portion of a 

facility subject to the early site permit, 

standard design certification, standard 

design approval, construction permit, 

combined license or manufacturing 

licensing requirements of parts 50, 52, 

or 53 of this chapter, provided the 

deviation could, on the basis of an 

evaluation, create a substantial safety 

hazard and the portion of the facility 

containing the deviation has been 

offered to the purchaser for acceptance; 

(4) A condition or circumstance 

involving a basic component that could 

contribute to the exceeding of a safety 

limit, as defined in the technical 

specifications of a license for operation 

issued under part 50, part 52, or part 53 

of this chapter; or 

(5) An error, omission or other 

circumstance in a design certification, 

or standard design approval that, on the 

basis of an evaluation, could create a 

substantial safety hazard. 

* * * * * 

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. 

* * * * * 

§ 21.21 [Amended] 

■ 70. Amend § 21.21 by: 

■ a. In paragraph (a)(3), removing the 

phrase ‘‘under part 52’’ and add in its 

place the phrase ‘‘under parts 52 or 53’’; 

and 

■ b. In paragraphs (d)(1)(i) and (ii) 

removing the phrase ‘‘parts 30, 40, 50, 

52, 60, 61, 63, 70, 71, or 72 of this 

chapter’’ and adding in its place the 

phrase ‘‘parts 30, 40, 50, 52, 53, 60, 61, 

63, 70, 71, or 72 of this chapter’’. 

§ 21.51 [Amended] 

■ 71. In § 21.51, in paragraphs (a)(4) and 

(5) remove the phrase ‘‘of part 52’’ and 

add in its place the phrase ‘‘of part 52 

or under subpart H of part 53’’. 

§ 21.61 [Amended] 

■ 72. In § 21.61, in paragraph (b) remove 

the phrase ‘‘under part 52’’ and add in 

its place the phrase ‘‘under parts 52 or 

53’’. 

PART 25—ACCESS AUTHORIZATION 

■ 73. The authority citation for part 25 

continues to read as follows: 

Authority: Atomic Energy Act of 1954, 

secs. 145, 161, 223, 234 (42 U.S.C. 2165, 

2201, 2273, 2282); Energy Reorganization Act 

of 1974, sec. 201 (42 U.S.C. 5841); 44 U.S.C. 

3504 note; E.O. 10865, 25 FR 1583, as 

amended, 3 CFR, 1959–1963 Comp., p. 398; 

E.O. 12829, 58 FR 3479, 3 CFR, 1993 Comp., 

p. 570; E.O. 13526, 75 FR 707, 3 CFR, 2009 

Comp., p. 298; E.O. 12968, 60 FR 40245, 3 

CFR, 1995 Comp., p. 391. Section 25.17(f) 

and Appendix A also issued under 31 U.S.C. 

9701; 42 U.S.C. 2214. 

■ 74. In § 25.5, revise the definition for 

‘‘License’’ to read as follows: 

§ 25.5 Definitions. 

* * * * * 

License means a license issued 

pursuant to 10 CFR parts 50, 52, 53, 60, 

63, 70, or 72. 

* * * * * 

§ 25.17 [Amended] 

■ 75. In § 25.17, in paragraph (a), 

remove the phrase ‘‘under 10 CFR parts 

50, 52, 54, 60, 63, 70, 72, or 76’’ and add 

in its place the phrase ‘‘under 10 CFR 

parts 50, 52, 53, 54, 60, 63, 70, 72, or 

76’’. 

§ 25.35 [Amended] 

■ 76. In § 25.35, in paragraph (a), 

wherever it appears, remove the phrase 

‘‘under part 52’’ and add in its place the 

phrase ‘‘under parts 52 or 53’’. 

PART 26—FITNESS FOR DUTY 

PROGRAMS 

■ 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. 

■ 78. In § 26.3, revise paragraph (d) and 

add paragraph (f) to read as follows: 

§ 26.3 Scope. 

* * * * * 

(d) Contractor/vendors (C/Vs) who 

implement FFD programs or program 

elements, to the extent that the licensees 

and other entities specified in 

paragraphs (a) through (c) and (f) of this 

section rely on those C/V FFD programs 

or program elements to meet the 

requirements of this part, shall comply 

with the requirements of this part. 

* * * * * 

(f) No later than the start of 

construction activities, licensees and 

other entities that have applied for or 

have been issued a license under part 53 

of this chapter, other than a 

manufacturing license (ML), must 

implement the requirements in subpart 

M of this part or all the requirements of 

this part except subparts K and M. 

Holders of an ML under part 53 of this 

chapter must implement the 

requirements in subpart M or all the 

requirements of this part except 

subparts K and M, before commencing 

activities that assemble a manufactured 

reactor. 

■ 79. In § 26.4, revise paragraphs (a) 

introductory text, (a)(1), (a)(4), (b), (c), 

(e) introductory text, (e)(4), (f), (g) 

introductory text, and (h) to read as 

follows: 

§ 26.4 FFD program applicability to 

categories of individuals. 

(a) All persons who are granted 

unescorted access to nuclear power 

reactor protected areas by the licensees 

in § 26.3(a) and, as applicable, (c) and 

perform the following duties shall be 

subject to an FFD program that meets all 

of the requirements of this part, except 

subpart K of this part, and those persons 

who are granted unescorted access to 

either nuclear power reactor protected 

areas or remote facilities where safetysignificant systems or components may 

be operated within the design basis of 


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a licensed commercial nuclear plant, by 

the licensees and other entities in 

§ 26.3(f) and perform the following 

duties must be subject to an FFD 

program that satisfies the requirements 

in subpart M of this part, unless the 

licensee or other entity subjects these 

individuals to an FFD program that 

satisfies all of the requirements of this 

part except for those requirements in 

subparts K and M: 

(1) For persons who are granted 

unescorted access by the licensees in 

§ 26.3(a) and, as applicable, (c), 

operating or onsite directing of the 

operation of systems and components 

that a risk-informed evaluation process 

has shown to be significant to public 

health and safety; for those persons who 

are granted unescorted access by the 

licensees and other entities in § 26.3(f), 

operating or directing of the operation of 

systems and components that a riskinformed evaluation process has shown 

to be significant to public health and 

safety; 

* * * * * 

(4) For persons who are granted 

unescorted access to nuclear power 

reactor protected areas by the licensees 

in § 26.3(a) and, as applicable, (c), 

performing maintenance or onsite 

directing of the maintenance of SSCs 

that a risk-informed evaluation process 

has shown to be significant to public 

health and safety; for those persons who 

are granted unescorted access to nuclear 

power reactor protected areas by the 

licensees and other entities in § 26.3(f), 

performing maintenance or directing of 

the maintenance of SSCs that a riskinformed evaluation process has shown 

to be significant to public health and 

safety; and 

* * * * * 

(b) All persons who are granted 

unescorted access to nuclear power 

reactor protected areas by the licensees 

in § 26.3(a) and, as applicable, (c) and 

who do not perform the duties 

described in paragraph (a) of this 

section shall be subject to an FFD 

program that meets all of the 

requirements of this part, except 

§§ 26.205 through 26.209 and subpart K 

of this part. All persons who are granted 

unescorted access to a facility licensed 

under part 53 of this chapter, and who 

do not perform or direct the 

performance of the duties described in 

§ 26.4(a), must be subject to the 

requirements in subpart M of this part, 

unless the licensee or other entity 

implements an FFD program that 

satisfies all of the requirements of this 

part, except §§ 26.205 through 26.209 

and subparts K and M. 

(c) All persons who are required by a 

licensee in § 26.3(a) and, as applicable, 

(c) to physically report to the licensee’s 

Technical Support Center or Emergency 

Operations Facility by licensee 

emergency plans and procedures shall 

be subject to an FFD program that meets 

all of the requirements of this part, 

except §§ 26.205 through 26.209 and 

subpart K of this part. Also, for licensees 

or other entities in § 26.3(f), all persons 

without unescorted access to the facility 

who make decisions and/or direct 

actions regarding plant safety and 

security, and all persons who 

participate remotely in emergency 

response activities or physically report 

to the Technical Support Center or 

Emergency Operations Facility (or an 

equivalent facility), must be subject to 

an FFD program that satisfies all of the 

requirements described in subpart M of 

this part, unless the licensee or other 

entity implements an FFD program that 

satisfies all of the requirements of this 

part, except §§ 26.205 through 26.209 

and subparts K and M. 

* * * * * 

(e) When construction activities, as 

defined in § 26.5, begin, any individual 

whose duties for the licensees and other 

entities in § 26.3(c) require him or her 

to have the following types of access or 

perform the following activities at the 

location where the nuclear power plant 

will be constructed and operated shall 

be subject to an FFD program that meets 

all of the requirements of this part, 

except subparts I, K, and M of this part, 

and for any individual whose duties for 

the licensees and other entities in 

§ 26.3(f) require him or her to have the 

following types of access, perform 

construction activities as defined in 

§ 26.5, or perform the following 

activities must be subject to an FFD 

program as described in subpart M or an 

FFD program that satisfies all of the 

requirements of this part, except 

subparts I, K, and M: 

* * * * * 

(4) Witnesses or determines 

inspections, tests, and analyses 

certification required under part 52 or 

part 53 of this chapter; 

* * * * * 

(f) Any individual who is constructing 

or directing the construction of safetyor security-related SSCs shall be subject 

to an FFD program that meets the 

requirements of subpart K, or, if 

applicable, subpart M of this part, 

unless the licensee or other entity 

subjects these individuals to an FFD 

program that meets all of the 

requirements of this part, except for 

subparts I, K, and M of this part. 

(g) All FFD program personnel who 

are involved in the day-to-day 

operations of the program, as defined by 

the procedures of the licensees and 

other entities in § 26.3(a) through (c), 

and, as applicable, (d) and whose duties 

require them to have the following types 

of access or perform the following 

activities shall be subject to an FFD 

program that meets all of the 

requirements of this part, except 

subparts I, K, and M of this part, and, 

at the licensee’s or other entity’s 

discretion, subpart C of this part. All 

personnel whose duties require them to 

have the following types of access or 

perform the following activities at 

facilities licensed under part 53 of this 

chapter must be subject to the 

requirements in subpart M or an FFD 

program that satisfies all of the 

requirements of this part, except 

subparts I, K, and M, and, at the 

licensee’s or other entity’s discretion, 

subpart C of this part: 

* * * * * 

(h) Individuals who have applied for 

authorization to have the types of access 

or perform the activities described in 

paragraphs (a) through (d) of this section 

shall be subject to §§ 26.31(c)(1), 

26.35(b), 26.37, 26.39, and the 

applicable requirements of subparts C, E 

through H, and M of this part. 

* * * * * 

■ 80. Amend § 26.5 by: 

■ a. Adding the definitions for 

‘‘Biological marker’’ and ‘‘Change’’; 

■ b. Revising the definitions for 

‘‘Constructing or construction 

activities’’ ‘‘Contractor/vendor (C/V)’’; 

■ c. Adding the definition of ‘‘Illicit 

substance’’; 

■ d. Revising the definitions of ‘‘Other 

entity’’ and ‘‘Questionable validity’’; 

■ e. Adding the definitions of 

‘‘Reduction in FFD program 

effectiveness’’; 

■ f. Revising the definitions of 

‘‘Reviewing official’’, ‘‘Safety-related 

structures, systems, and components 

(SSCs)’’, and ‘‘Security-related SSCs’’; 

■ g. Adding the definitions of ‘‘Special 

nuclear material’’; and 

■ h. Revising the definition of ‘‘Unit 

outage’’. 

The additions and revisions read as 

follows: 

§ 26.5 Definitions. 

* * * * * 

Biological marker means, for a part 53 

licensee implementing subpart M of this 

part, an endogenous substance that is 

used to validate that the biological 

specimen collected for testing was 

produced by the donor. 

* * * * * 


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Change as used in § 26.603(e) means 

an action that results in a modification 

of, addition to, or removal from the 

licensee’s or other entity’s FFD program. 

* * * * * 

Constructing or construction activities 

means, for the purposes of this part, the 

tasks involved in building a nuclear 

power plant that are performed at the 

location where the nuclear power plant 

will be constructed and operated. These 

tasks include fabricating, erecting, 

integrating, and testing safety- and 

security-related SSCs, and the 

installation of their foundations, 

including the placement of concrete. For 

a licensee or other entity described in 

§ 26.3(f), construction is defined in 

§ 53.020 of this chapter. 

Contractor/vendor (C/V) means any 

company, or any individual not 

employed by a licensee or other entity 

specified in § 26.3(a) through (c) and (f), 

who is providing work or services to a 

licensee or other entity covered in 

§ 26.3(a) through (c) and (f), either by 

contract, purchase order, oral 

agreement, or other arrangement. 

* * * * * 

Illicit substance means a substance 

that causes impairment and possible 

addiction but is not an illegal drug as 

defined in § 26.5. 

* * * * * 

Other entity means any corporation, 

firm, partnership, limited liability 

company, association, C/V, or other 

organization who is subject to this part 

under § 26.3(a) through (c) and (f) but is 

not licensed by the NRC. 

* * * * * 

Questionable validity means the 

results of validity screening or initial 

validity tests at a licensee testing facility 

indicating that a urine specimen may be 

adulterated, substituted, dilute, or 

invalid. For a part 53 licensee or other 

entity, questionable validity means the 

results of validity screening or initial 

validity tests indicating that a biological 

specimen obtained from an individual 

pursuant to subpart M of this part may 

be adulterated, substituted, dilute, or 

invalid. 

Reduction in FFD program 

effectiveness means, for a part 53 

licensee or other entity implementing 

subpart M of this part, a change or series 

of changes to an element of the FFD 

program that reduces or eliminates the 

licensee’s ability to satisfy or maintain 

site-specific FFD program performance 

when compared to historical sitespecific performance, the licensee’s 

fleet-level program performance, or 

industry performance. 

* * * * * 

Reviewing official means an employee 

of a licensee or other entity specified in 

§ 26.3(a) through (c) and (f), who is 

designated by the licensee or other 

entity to be responsible for reviewing 

and evaluating any potentially 

disqualifying FFD information about an 

individual, including, but not limited 

to, the results of a determination of 

fitness, as defined in § 26.189, in order 

to determine whether the individual 

may be granted or maintain 

authorization. 

Safety-related structures, systems, and 

components (SSCs) means, for part 50 or 

part 52 licensees and other entities 

described in § 26.3(a) through (d), those 

SSCs that are relied on to remain 

functional during and following design 

basis events to ensure the integrity of 

the reactor coolant pressure boundary, 

the capability to shut down the reactor 

and maintain it in a safe shutdown 

condition, or the capability to prevent or 

mitigate the consequences of accidents 

that could result in potential offsite 

exposure comparable to the guidelines 

in § 50.34(a)(1) of this chapter. For part 

53 licensees and other entities described 

in § 26.3(d) and (f), safety-related has 

the same meaning as that in § 53.020 of 

this chapter. 

Security-related SSCs means, for the 

purposes of this part, those structures, 

systems, and components that the 

licensee will rely on to implement the 

licensee’s physical security and 

safeguards contingency plans that either 

are required under part 73 of this 

chapter if the licensee is a construction 

permit applicant or holder or an early 

site permit holder, as described in 

§ 26.3(c)(3) through (c)(5), respectively, 

or are included in the licensee’s 

application if the licensee is a combined 

license applicant or holder, as described 

in § 26.3(c)(1) and (c)(2), respectively, or 

a licensee or other entity described in 

§ 26.3(d) or (f). 

* * * * * 

Special nuclear material (SNM) has 

the same meaning as that in § 70.4 of 

this chapter. 

* * * * * 

Unit outage means, for the purposes 

of this part, for electricity-generation 

units, that the reactor unit is 

disconnected from the electrical grid. 

Unit outage means, for the purposes of 

this part, for non-electricity-generation 

units, that the reactor unit is 

disconnected from the loads to which 

its output is supplied under normal 

operating conditions. 

* * * * * 

■ 81. In § 26.8, revise paragraph (b) to 

read as follows: 

§ 26.8 Information collection 

requirements: OMB approval. 

* * * * * 

(b) The approved information 

collection requirements contained in 

this part appear in §§ 26.9, 26.27, 26.29, 

26.31, 26.33, 26.35, 26.37, 26.39, 26.41, 

26.53, 26.55, 26.57, 26.59, 26.61, 26.63, 

26.65, 26.67, 26.69, 26.75, 26.77, 26.85, 

26.87, 26.89, 26.91, 26.93, 26.95, 26.97, 

26.99, 26.101, 26.103, 26.107, 26.109, 

26.111, 26.113, 26.115, 26.117, 26.119, 

26.125, 26.127, 26.129, 26.135, 26.137, 

26.139, 26.153, 26.157, 26.159, 26.163, 

26.165, 26.167, 26.168, 26.169, 26.183, 

26.185, 26.187, 26.189, 26.202, 26.203, 

26.205, 26.207, 26.211, 26.401, 26.403, 

26.405, 26.406, 26.407, 26.411, 26.413, 

26.415, 26.417, 26.603, 26.604, 26.605, 

26.606, 26.607, 26.608, 26.609, 26.611, 

26.613, 26.617, 26.619, 26.711, 26.713, 

26.715, 26.717, 26.719, and 26.821. 

■ 82. Revise § 26.21 to read as follows: 

§ 26.21 Fitness-for-duty program. 

The licensees and other entities 

specified in § 26.3(a) through (c) and (f) 

(for those licensees and other entities 

that do not implement the requirements 

in subparts M and K of this part) shall 

establish, implement, and maintain FFD 

programs that, at a minimum, comprise 

the program elements contained in this 

subpart. The individuals specified in 

§ 26.4(a) through (e) and (g), and, at the 

licensee’s or other entity’s discretion, 

§ 26.4(f), and, if necessary, § 26.4(j) shall 

be subject to these FFD programs. 

Licensees and other entities may rely on 

the FFD program or program elements of 

a C/V, as defined in § 26.5, if the C/V’s 

FFD program or program elements 

satisfy the applicable requirements of 

this part. 

■ 83. Revise § 26.51 to read as follows: 

§ 26.51 Applicability. 

The requirements in this subpart 

apply to the licensees and other entities 

identified in § 26.3(a), (b), and, as 

applicable, (c) for the categories of 

individuals in § 26.4(a) through (d), and, 

at the licensee’s or other entity’s 

discretion, in § 26.4(g) and, if necessary, 

§ 26.4(j). The requirements in this 

subpart also apply to the licensees and 

other entities specified in § 26.3(c), as 

applicable, for the categories of 

individuals in § 26.4(e). At the 

discretion of a licensee or other entity 

in § 26.3(c), the requirements of this 

subpart also may be applied to the 

categories of individuals identified in 

§ 26.4(f). In addition, the requirements 

in this subpart apply to the entities in 

§ 26.3(d) to the extent that a licensee or 

other entity relies on the C/V to satisfy 

the requirements of this subpart. Certain 

requirements in this subpart also apply 


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to the individuals specified in § 26.4(h). 

The requirements in this subpart apply 

to the FFD programs of licensees and 

other entities identified in § 26.3(f) that 

elect not to implement the requirements 

in subpart M for the categories of 

individuals in § 26.4 and those licensees 

and other entities that elect to 

implement the requirements in § 26.605. 

§ 26.53 [Amended] 

■ 84. Amend § 26.53 by: 

■ a. In paragraph (e), wherever it 

appears, remove the phrase ‘‘§ 26.3(a) 

through (c)’’ and add in its place the 

phrase ‘‘§ 26.3(a) through (c) and (f)’’; 

and 

■ b. In paragraphs (g), (h), and (i), 

wherever it appears, remove the phrase 

‘‘(c) and (d)’’ and add in its place the 

phrase ‘‘(c), (d), and (f)’’. 

§ 26.63 [Amended] 

■ 85. In § 26.63, in paragraph (d) remove 

the phrase ‘‘§ 26.3(a) through (d)’’ and 

add in its place the phrase ‘‘§ 26.3(a) 

through (d) and (f)’’. 

■ 86. Revise § 26.73 to read as follows: 

§ 26.73 Applicability. 

The requirements in this subpart 

apply to the licensees and other entities 

identified in § 26.3(a), (b), and, as 

applicable, (c) for the categories of 

individuals specified in § 26.4(a) 

through (d) and (g). The requirements in 

this subpart also apply to the licensees 

and other entities specified in § 26.3(c), 

as applicable, for the categories of 

individuals in § 26.4(e). At the 

discretion of a licensee or other entity 

in § 26.3(c), the requirements of this 

subpart also may be applied to the 

categories of individuals identified in 

§ 26.4(f). In addition, the requirements 

in this subpart apply to the entities in 

§ 26.3(d) to the extent that a licensee or 

other entity relies on the C/V to satisfy 

the requirements of this subpart. The 

regulations in this subpart also apply to 

the individuals specified in § 26.4(h) 

and (j), as appropriate. The 

requirements in this subpart apply to 

the FFD programs of licensees and other 

entities identified in § 26.3(f) that elect 

not to implement the requirements in 

subpart M for the categories of 

individuals in § 26.4 and those licensees 

and other entities that elect to 

implement the requirements in 

§ 26.605(b). 

■ 87. Revise § 26.81 to read as follows: 

§ 26.81 Purpose and applicability. 

This subpart contains requirements 

for collecting specimens for drug testing 

and conducting alcohol tests by or on 

behalf of the licensees and other entities 

in § 26.3(a) through (d) for the categories 

of individuals specified in § 26.4(a) 

through (d) and (g). At the discretion of 

a licensee or other entity in § 26.3(c), 

specimen collections and alcohol tests 

must be conducted either under this 

subpart for the individuals specified in 

§ 26.4(e) and (f) or the licensee or other 

entity may rely on specimen collections 

and alcohol tests conducted under the 

requirements of 49 CFR part 40 for the 

individuals specified in § 26.4(e) and (f). 

The requirements of this subpart do not 

apply to specimen collections and 

alcohol tests that are conducted under 

the requirements of 49 CFR part 40, as 

permitted in this paragraph and under 

§§ 26.4(j) and 26.31(b)(2) and subpart K. 

The requirements in this subpart apply 

to the FFD programs of licensees and 

other entities identified in § 26.3(f) that 

elect not to implement the requirements 

in subpart M for the categories of 

individuals in § 26.4 and those licensees 

and other entities that elect to 

implement the requirements in § 26.605. 

■ 88. Revise § 26.201 to read as follows: 

§ 26.201 Applicability. 

(a) The requirements in this subpart, 

with the exception of § 26.202, apply to 

the licensees and other entities 

identified in § 26.3(a); if applicable, (c), 

(d), and (f), for licensees and other 

entities not implementing the 

requirements in subparts K and M. For 

the licensees and other entities to whom 

the requirements in this subpart, with 

the exception of § 26.202, apply, the 

requirements in §§ 26.203 and 26.211 

apply to the individuals identified in 

§ 26.4(a) through (c). In addition, the 

requirements in §§ 26.205 through 

26.209 apply to the individuals 

identified in § 26.4(a). 

(b) The requirements in this subpart, 

with the exception of § 26.203, apply to 

the licensees or other entities identified 

in § 26.3(f) implementing this subpart 

under §§ 26.604 and 26.605. For these 

licensees and other entities, the 

requirements in §§ 26.202 and 26.211 

apply to the individuals identified in 

§ 26.4(a) through (c) and any person 

licensed to operate under 10 CFR part 

53; and the requirements in §§ 26.205 

through 26.209 apply to the individuals 

identified in § 26.4(a). 

■ 89. Add § 26.202 to read as follows: 

§ 26.202 General provisions for facilities 

licensed under part 53. 

(a) Policy. Licensees must establish a 

policy for the management of fatigue for 

all individuals who are subject to the 

licensee’s FFD program and incorporate 

it into the written policy required in 

§ 26.606(a). 

(b) Procedures. In addition to the 

procedures required in § 26.606(b), 

licensees must develop, implement, and 

maintain procedures that— 

(1) Describe the process to be 

followed when any individual 

identified in § 26.4(a) through (c) makes 

a self-declaration that he or she is not 

fit to safely and competently perform 

his or her duties for any part of a 

working tour as a result of fatigue. The 

procedure must— 

(i) Describe the individual’s and 

licensee’s rights and responsibilities 

related to self-declaration; 

(ii) Describe requirements for 

establishing controls and conditions 

under which an individual may be 

permitted or required to perform work 

after that individual declares that he or 

she is not fit due to fatigue; and 

(iii) Describe the process to be 

followed if the individual disagrees 

with the results of a fatigue assessment 

that is required under § 26.211(a)(2); 

(2) Describe the process for 

implementing the controls required 

under § 26.205 for the individuals who 

are performing the duties listed in 

§ 26.4(a); 

(3) Describe the process to be 

followed in conducting fatigue 

assessments under § 26.211; and 

(4) Describe the disciplinary actions 

that the licensee may impose on an 

individual following a fatigue 

assessment, and the conditions and 

considerations for taking those 

disciplinary actions. 

(c) Training and assessments. 

Licensees must include the following 

KAs in the content of the training and 

trainee assessments required in 

§ 26.608: 

(1) Knowledge of the contributors to 

worker fatigue, circadian variations in 

alertness and performance, indications 

and risk factors for common sleep 

disorders, shiftwork strategies for 

obtaining adequate rest, and the 

effective use of fatigue countermeasures; 

and 

(2) Ability to identify symptoms of 

worker fatigue and contributors to 

decreased alertness in the workplace. 

(d) Recordkeeping. Licensees must 

retain the following records for at least 

3 years or until the completion of all 

related legal proceedings, whichever is 

later: 

(1) Records of work hours for 

individuals who are subject to the work 

hour controls in § 26.205; 

(2) For licensees implementing the 

requirements of § 26.205(d)(3), records 

of shift schedules and shift cycles, or, 

for licensees implementing the 

requirements of § 26.205(d)(7), records 

of shift schedules and records showing 

the beginning and end times and dates 

of all averaging periods, of individuals 


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who are subject to the work hour 

controls in § 26.205; 

(3) The documentation of waivers that 

is required in § 26.207(a)(4), including 

the bases for granting the waivers; 

(4) The documentation of work hour 

reviews that is required in § 26.205(e)(3) 

and (e)(4); and 

(5) The documentation of fatigue 

assessments that is required in 

§ 26.211(g). 

(e) Reporting. Licensees must include 

the following information in a standard 

format in the annual FFD program 

performance report required under 

§ 26.617(b)(2): 

(1) A summary for each nuclear power 

plant site of all instances during the 

previous calendar year when the 

licensee waived one or more of the work 

hour controls specified in § 26.205(d)(1) 

through (d)(5)(i) and (d)(7) for 

individuals described in § 26.4(a). The 

summary must include only those 

waivers under which work was 

performed. If it was necessary to waive 

more than one work hour control during 

any single extended work period, the 

summary of instances must include 

each of the work hour controls that were 

waived during the period. For each 

category of individuals specified in 

§ 26.4(a), the licensee must report— 

(i) The number of instances when 

each applicable work hour control 

specified in § 26.205(d)(1)(i) through 

(iii), (d)(2)(i) and (ii), (d)(3)(i) through 

(v), and (d)(7) was waived for 

individuals not working on outage 

activities; 

(ii) The number of instances when 

each applicable work hour control 

specified in § 26.205(d)(1)(i) through 

(iii), (d)(2)(i) and (ii), (d)(3)(i) through 

(v), (d)(4) and (d)(5)(i), and (d)(7) was 

waived for individuals working on 

outage activities; and 

(iii) A summary that shows the 

distribution of waiver use among the 

individuals applicable within each 

category of individuals identified in 

§ 26.4(a) (e.g., a table that shows the 

number of individuals who received 

only one waiver during the reporting 

period, the number of individuals who 

received a total of two waivers during 

the reporting period). 

(2) A summary of corrective actions, 

if any, resulting from the analyses of 

these data, including fatigue 

assessments. 

(f) Audits. Licensees must audit the 

management of worker fatigue under 

§ 26.615. 

■ 90. In § 26.205, revise paragraphs 

(d)(7)(iii) and (d)(8) to read as follows: 

§ 26.205 Work Hours. 

* * * * * 

(d) * * * 

(7) * * * 

(iii) Each licensee shall state, in its 

FFD policy and procedures required by 

either §§ 26.27 and 26.203(a) and (b) or 

§§ 26.202(a) and (b) and 26.606, the 

work hour counting system in 

§ 26.205(d)(7)(ii) the licensee is using. 

(8) Each licensee shall state, in its 

FFD policy and procedures required by 

either §§ 26.27 and 26.203(a) and (b) or 

§§ 26.202(a) and (b) and 26.606, the 

requirements with which the licensee is 

complying: the minimum days off 

requirements in § 26.205(d)(3) or 

maximum average work hours 

requirements in § 26.205(d)(7). 

* * * * * 

■ 91. In § 26.207, revise paragraph 

(a)(1)(ii) to read as follows: 

§ 26.207 Waivers and exceptions. 

(a) * * * 

(1) * * * 

(ii) A supervisor assesses the 

individual face to face and determines 

that there is reasonable assurance that 

the individual will be able to safely and 

competently perform his or her duties 

during the additional work period for 

which the waiver will be granted. The 

supervisor performing the assessment 

shall be trained as required by either 

§§ 26.29 and 26.203(c) or §§ 26.202(c) 

and 26.608 and shall be qualified to 

direct the work to be performed by the 

individual. If there is no supervisor on 

site who is qualified to direct the work, 

the assessment may be performed by a 

supervisor who is qualified to provide 

oversight of the work to be performed by 

the individual. At a minimum, the 

assessment must address the potential 

for acute and cumulative fatigue 

considering the individual’s work 

history for at least the past 14 days, the 

potential for circadian degradations in 

alertness and performance considering 

the time of day for which the waiver 

will be granted, the potential for fatiguerelated degradations in alertness and 

performance to affect risk-significant 

functions, and whether any controls and 

conditions must be established under 

which the individual will be permitted 

to perform work. For licensees and other 

entities in § 26.3(f), the assessment may 

be performed remotely using electronic 

communications. In such instances, the 

assessment must be supported by 

someone who is present in-person with 

the individual whose alertness may be 

impaired, and that supporting person 

must be trained under the requirements 

of either § 26.29 and § 26.203(c) or 

§ 26.202(c) and § 26.608. 

* * * * * 

■ 92. In § 26.211, revise paragraphs 

(a)(1) and (3) and paragraph (b) 

introductory text to read as follows: 

§ 26.211 Fatigue assessments. 

(a) * * * 

(1) For-cause. In addition to any other 

test or determination of fitness that may 

be required under §§ 26.31(c), 26.77, 

26.607(b), and 26.619, a fatigue 

assessment must be conducted in 

response to an observed condition of 

impaired individual alertness creating a 

reasonable suspicion that an individual 

is not fit to safely and competently 

perform his or her duties, except if the 

condition is observed during an 

individual’s break period. If the 

observed condition is impaired alertness 

with no other behaviors or physical 

conditions creating a reasonable 

suspicion of possible substance abuse, 

then the licensee need only conduct a 

fatigue assessment. If the licensee has 

reason to believe that the observed 

condition is not due to fatigue, the 

licensee need not conduct a fatigue 

assessment; 

* * * * * 

(3) Post-event. A fatigue assessment 

must be conducted in response to events 

requiring post-event drug and alcohol 

testing as specified in § 26.31(c) or postevent tests in § 26.607(b)(4). Licensees 

may not delay necessary medical 

treatment in order to conduct a fatigue 

assessment; and 

* * * * * 

(b) Only supervisors and FFD program 

personnel who are trained under either 

§§ 26.29 and 26.203(c) or §§ 26.202(c) 

and 26.608 may conduct a fatigue 

assessment. The fatigue assessment 

must be conducted face to face with the 

individual whose alertness may be 

impaired. For licensees and other 

entities in § 26.3(f), a fatigue assessment 

may be performed remotely using 

electronic communications. In such 

instances, the fatigue assessment must 

be supported by someone who is 

present in-person with the individual 

whose alertness may be impaired, and 

that supporting person must be trained 

in accordance with the requirements of 

either §§ 26.29 and 26.203(c) or 

§§ 26.202(c) and 26.608. 

* * * * * 

■ 93. Add Subpart M, consisting of 

§§ 26.601 through 26.619, to read as 

follows: 

Subpart M—Fitness for Duty Programs 

for Facilities Licensed Under 10 CFR 

Part 53 

Sec. 

26.601 Applicability. 

26.603 General provisions. 


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26.604 FFD program requirements for 

facilities that satisfy the § 26.603(c) 

criterion. 

26.605 FFD program requirements for 

facilities that do not implement § 26.604. 

26.606 Written policy and procedures. 

26.607 Drug and alcohol testing. 

26.608 FFD program training. 

26.609 Behavioral observation. 

26.610 Sanctions. 

26.611 Protection of information. 

26.613 Appeals process. 

26.615 Audits. 

26.617 Recordkeeping and reporting. 

26.619 Suitability and fitness 

determinations. 

§ 26.601 Applicability. 

A licensee or other entity in § 26.3(f), 

at its discretion, may establish, 

implement, and maintain a fitness-forduty (FFD) program that satisfies the 

requirements of this subpart for those 

categories of individuals in § 26.4, as 

applicable, and any person licensed to 

operate under 10 CFR part 53. If a 

licensee or other entity in § 26.3(f) does 

not elect to implement an FFD program 

that satisfies the requirements of this 

subpart, then those categories of 

individuals in § 26.4, as applicable, and 

any person licensed to operate under 10 

CFR part 53 must be subject to an FFD 

program that satisfies all part 26 

requirements, except for those 

requirements in subparts K and M. 

§ 26.603 General provisions. 

(a) FFD program description. An 

applicant’s description of the FFD 

program in its Final Safety Analysis 

Report, required by subpart H of part 53 

of this chapter, must include— 

(1) If the applicant performed the 

analysis under paragraph (c) of this 

section, a summary of the analysis, 

including the assumptions, 

methodology, conclusion, and 

references; 

(2) A statement whether the FFD 

program will be implemented pursuant 

to § 26.604 or § 26.605, or will satisfy all 

part 26 requirements, except for the 

requirements in subparts K and M; 

(3) A discussion of the applicability of 

the FFD program to those individuals 

described in § 26.4 and how the 

program will be implemented offsite at 

a U.S. Nuclear Regulatory Commission 

(NRC)-licensed facility authorized to 

assemble or test a manufactured reactor, 

if applicable; 

(4) A description of the drug and 

alcohol testing and fitness 

determination process to be 

implemented through the licensee’s or 

other entity’s procedures, including the 

collection and testing facilities to be 

used, biological specimens to be 

collected, and sanctions to be imposed 

upon a confirmed FFD policy violation; 

and 

(5) A summary of the FFD 

performance monitoring and review 

program (PMRP), including the 

measures and thresholds required by 

paragraph (d)(1) of this section. 

(b) FFD program implementation and 

availability. For the licensees and other 

entities in § 26.3(f), other than the 

holder of a manufacturing license (ML), 

the FFD program must be implemented 

no later than the start of construction 

activities, as defined in § 26.5, and 

maintained until the NRC’s docketing of 

the license holder’s certifications 

described in § 53.1070 of this chapter. 

For holders of an ML, the FFD program 

must be implemented no later than the 

start of activities that assemble the 

manufactured reactor and maintained 

until expiration of the ML. 

(c) Criterion and analysis for an FFD 

program. For a licensee or other entity 

to implement an FFD program under 

§ 26.604, the licensee or other entity 

must perform a site-specific analysis to 

demonstrate that the facility and its 

operation satisfy the criterion in 

§ 53.860(a)(2) of this chapter. The 

licensee or other entity must maintain 

the analysis, including updates to reflect 

changes made to the staffing, FFD 

programs, or offsite support resources 

described in the analysis, to show that 

the facility and its operation continue to 

satisfy the criterion, until permanent 

cessation of operations under § 53.1070 

of this chapter. 

(d) FFD performance monitoring and 

review. A licensee or other entity must 

establish performance measures and 

associated thresholds as described in 

paragraph (d)(1) of this section and 

monitor the effectiveness of its FFD 

program by comparing performance data 

against these performance measures and 

thresholds, in a manner sufficient to 

satisfy the § 26.23 performance 

objectives. 

(1) PRMP elements. The PMRP must 

be documented and maintained and 

include the following program elements: 

(i) Performance measures. 

Performance measures must be 

identified and designed to monitor FFD 

program performance. 

(A) If the licensee or other entity is 

subject to the requirements in § 26.604, 

then the monitoring program must 

include performance measures for the 

following: the behavioral observation 

program; occurrence of FFD policy 

violations categorized by licensee 

employee, contractor/vendor, and labor 

category; and occurrence of individuals 

with potentially disqualifying 

information or who possessed FFD 

prohibited items. 

(B) If the licensee or other entity is 

subject to the requirements in § 26.604 

and has implemented a drug testing 

program at its discretion or is subject to 

the requirements of § 26.605, then the 

monitoring program must include 

performance measures identified in 

paragraph (d)(1)(i)(A) of this section. 

This monitoring program must also 

include performance measures for the 

pre-access and random positive testing 

rates, random testing rate for licensee 

employees and contractor/vendors, and 

the number of subversion attempts 

categorized by licensee employee, 

contractor/vendor, and labor category. 

(ii) Thresholds. Licensee- or other 

entity-specific thresholds for its sitespecific performance measures must be 

established and used to facilitate 

corrective actions to maintain FFD 

program performance. Initial thresholds 

must be based on FFD performance data 

from comparable facilities subject to 

part 26, the licensee’s or other entity’s 

fleet-level program performance if 

applicable, and industry FFD 

performance data. 

(iii) Monitoring program. Licensees 

and other entities must monitor the 

performance of their FFD programs 

against licensee- or other entityestablished performance measures and 

thresholds as FFD performance data is 

received to determine whether a 

threshold has been exceeded. Licensees 

and other entities must perform year-toyear comparisons of site-specific 

performance; site-specific performance 

to the licensee’s or other entity’s fleetlevel program performance, if 

applicable; and site-specific to industry 

performance. 

(iv) Quantitative and qualitative 

reviews. The PMRP must include a 

documented review of the elements in 

paragraph (d)(1)(i) through (iii) of this 

section and the following qualitative 

elements. 

(A) Worker protections. The review 

must include a documented assessment 

of the licensee’s or other entity’s 

implementation of the protections 

described in §§ 26.606(b)(1)(iii), 26.611, 

and 26.613. 

(B) Laboratory test results and 

Medical Review Officer performance. 

The review must include a documented 

assessment of whether the actions taken 

by the Medical Review Officer (MRO) 

met the requirements in § 26.185 based 

on the laboratory test results reported 

under § 26.169. This review must 

include a comparative analysis between 

the point of collection testing and 

assessment (POCTA) screening result(s) 

and the corresponding specimen test 

results obtained from the U.S. 

Department of Health and Human 


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Services (HHS)-certified laboratory if 

the POCTA indicated a positive, 

adulterated, substituted, or invalid 

screening result or discrepant biological 

marker, to assess the effectiveness of the 

POCTA and to inform MRO decisions 

under § 26.185 or § 26.607(m)(6). 

(C) Change control. The review must 

include a documented assessment of the 

changes made under paragraph (e) of 

this section to verify that the summation 

of program changes has not resulted in 

a reduction in FFD program 

effectiveness. 

(2) Corrective actions. Corrective 

actions must be implemented to address 

when FFD performance meets a 

licensee-established performance 

threshold or to resolve a finding 

resulting from a qualitative review or 

audit in a manner that restores 

performance and corrects root causes, 

contributing causes, or both. 

(3) Program review periodicity. The 

documented review in paragraph 

(d)(1)(iv) of this section must be 

conducted biennially to assess and 

modify licensee or other entity 

implementation of its FFD program. 

This documented review must 

demonstrate that the performance 

measures and thresholds are appropriate 

and adjusted as necessary based on sitelevel and licensee’s or other entity’s 

fleet-level, if applicable, program 

performance, and industry performance. 

(i) Identified program weaknesses and 

corrective actions must be summarized 

in the annual reporting requirement 

described in § 26.617(b)(2) or § 26.717, 

as applicable. 

(ii) The program review must be 

completed and approved by the licensee 

or other entity to support the reporting 

of PMRP weaknesses and corrective 

actions as required in paragraph (d)(3)(i) 

of this section every odd-numbered 

year, and the implementation of 

corrective actions before May 15 of that 

odd-numbered year. 

(e) FFD program change control. (1) 

The licensee or other entity may make 

changes to its FFD program under this 

subpart if— 

(i) The licensee or other entity 

performs and retains an analysis 

demonstrating that the changes do not 

reduce the effectiveness of the FFD 

program; or 

(ii) The change was necessitated or 

justified by a change to part 26, 

laboratory processes or procedures, or 

guidance issued by the HHS or NRC, as 

implemented by the licensee or other 

entity though its procedures. 

(2) A licensee or other entity desiring 

to make a change that decreases FFD 

program effectiveness must implement a 

mitigating strategy so the FFD program, 

as revised, will continue to satisfy the 

performance objectives in § 26.23 and 

not result in a reduction in program 

effectiveness. 

(3) Except for phencyclidine, and 

notwithstanding paragraph (e)(1)(ii) of 

this section, the change control process 

may not be used to reduce the minimum 

panel of drugs to be tested in 

§ 26.607(c)(1). 

(4) The licensee must retain a record 

of each change made under this section 

for a period of at least 5 years from the 

date the change was implemented and 

summarize this change in its annual 

FFD performance report required by 

§ 26.617(b)(2) or § 26.717, as applicable. 

§ 26.604 FFD program requirements for 

facilities that satisfy the § 26.603(c) 

criterion. 

(a) FFD program. A licensee or other 

entity with an analysis that 

demonstrates that its facility and 

operation satisfy the criterion in 

§ 26.603(c) may elect to establish, 

implement, and maintain an FFD 

program under this section. That FFD 

program must contain the following 

elements: 

(1) Applies to those individuals 

described in § 26.4, as applicable; and 

(2) Implements the following 

requirements and subparts in this part: 

(i) § 26.23, Performance objectives; 

(ii) § 26.603, General provisions; 

(iii) § 26.606, Written policies and 

procedures, (a) and, if applicable (b); 

(iv) § 26.608, FFD program training; 

(v) § 26.609, Behavioral observation; 

(vi) § 26.610, Sanctions; 

(vii) § 26.611, Protection of 

information; 

(viii) § 26.613, Appeals process; 

(ix) § 26.615, Audits; 

(x) § 26.617, Recordkeeping and 

reporting; 

(xi) § 26.619, Suitability and fitness 

determinations; 

(xii) Subpart A—Administrative 

Provisions; 

(xiii) Subpart I—Managing Fatigue; 

and 

(xiv) Subpart O—Inspections, 

Violations, and Penalties. 

(b) [Reserved] 

§ 26.605 FFD program requirements for 

facilities that do not implement § 26.604. 

(a) Licensees and other entities who 

satisfy the criterion in § 26.603(c), at 

their discretion, and licensees and other 

entities who do not satisfy the criterion 

in § 26.603(c), must establish, 

implement, and maintain an FFD 

program under this section either during 

construction activities as defined in 

§ 26.5, or during activities performed 

under an ML that allows the assembly, 

testing, or both of a manufactured 

reactor, as applicable. This FFD program 

must contain the following elements: 

(1) Applies to those individuals 

described in § 26.4, as applicable; and, 

(2) Implements the following 

requirements and subparts in this part— 

(i) § 26.23, Performance objectives; 

(ii) § 26.603, General provisions; 

(iii) § 26.606, Written policy and 

procedures; 

(iv) § 26.607, Drug and alcohol testing; 

(v) § 26.608, FFD program training; 

(vi) § 26.609, Behavioral observation; 

(vii) § 26.610, Sanctions; 

(viii) § 26.611, Protection of 

information; 

(ix) § 26.613, Appeals process; 

(x) § 26.615, Audits; 

(xi) § 26.617, Recordkeeping and 

reporting; 

(xii) § 26.619, Suitability and fitness 

determinations; 

(xiii) Subpart A—Administrative 

Provisions; 

(xiv) Subpart I—Managing Fatigue, in 

the case of holders of an ML that allows 

the assembly, testing, or both of a 

manufactured reactor; and 

(xv) Subpart O—Inspections, 

Violations, and Penalties. 

(b) Licensees and other entities who 

satisfy the criterion in § 26.603(c), at 

their discretion, and licensees and other 

entities who do not satisfy the criterion 

in § 26.603(c), before the loading of fuel 

onsite into a reactor vessel; before 

receiving a manufactured reactor; or 

before individuals subject to part 26 

operate, test, perform maintenance of, or 

direct the maintenance or surveillance 

of security-related equipment or 

equipment that a risk-informed 

evaluation process has shown to be 

significant to public health and safety, 

must establish, implement, and 

maintain an FFD program that— 

(1) Applies to those individuals 

described in § 26.4, as applicable; and, 

(2) Implements the following 

requirements and subparts— 

(i) § 26.23, Performance objectives; 

(ii) § 26.603, General provisions; 

(iii) § 26.606, Written policy and 

procedures; 

(iv) § 26.607, Drug and alcohol testing; 

(v) § 26.608, FFD program training; 

(vi) § 26.609, Behavioral observation; 

(vii) § 26.611, Protection of 

information; 

(viii) § 26.613, Appeals process; 

(ix) § 26.615, Audits; 

(x) Subpart A—Administrative 

Provisions; 

(xi) Subpart C—Granting and 

Maintaining Authorization; 

(xii) Subpart D—Management Actions 

and Sanctions to be Imposed; 

(xiii) Subpart H—Determining 

Fitness-for-Duty Policy Violations and 


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Determining Fitness, unless using the 

HHS Guidelines for MRO evaluation of 

drug test results, and determining 

fitness; 

(xiv) Subpart I—Managing Fatigue; 

(xv) Subpart N—Recordkeeping and 

Reporting Requirements; and 

(xvi) Subpart O—Inspections, 

Violations, and Penalties. 

§ 26.606 Written policy and procedures. 

(a) Licensees and other entities that 

implement an FFD program under this 

subpart must ensure that— 

(1) A written FFD policy statement is 

provided to each individual who is 

subject to the program before the 

individual is subject to behavioral 

observation, drug and alcohol testing, or 

both. 

(2) The FFD policy statement 

describes the performance objectives in 

§ 26.23. 

(3) The FFD policy statement 

describes the minimum days off 

requirements in § 26.205(d)(3) or 

maximum average work hours 

requirements in § 26.205(d)(7). 

(4) The FFD policy statement must be 

written in sufficient detail to provide 

affected individuals with information 

on what is expected of them and what 

consequences may result from a lack of 

adherence to the policy, including those 

elements described in § 26.606(b), part 

26-required sanctions, and required 

medical/clinical treatment and followup testing for FFD policy violations. 

(5) The FFD policy statement 

describes the individual’s 

responsibilities to report for work in a 

physiological and psychological 

condition that enables the safe and 

competent performance of assigned 

duties and responsibilities and inform a 

licensee- or other entity-designated 

representative when the individual 

determines that this cannot be 

accomplished. 

(b) Licensees and other entities must 

establish, implement, and maintain 

written procedures that address the 

following topics: 

(1) If implementing a drug and 

alcohol testing program under this 

subpart, 

(i) The methods and techniques to 

collect and test for drugs and alcohol 

and for the shipping and temporary 

storage of biological specimens used for 

drug testing at HHS-certified 

laboratories, 

(ii) The urine specimen volumes, 

techniques for split specimen 

collections, and the acceptability of a 

urine specimen as described in § 26.111 

or as described in the HHS Guidelines, 

(iii) Protecting the privacy of an 

individual who provides a specimen, 

protecting the integrity of the specimen, 

and ensuring that the test results are 

valid and attributable to the correct 

individual, and 

(iv) If the licensee or other entity 

elects to use the HHS Guidelines, the 

name of the specific HHS Guideline and 

revision being implemented by the 

licensee or other entity and a 

description of the specific sections in 

the guideline that are being 

implemented in the procedure, 

including specimen collections, drug 

testing, and evaluation of test results. 

(2) The immediate and follow-up 

actions that will be taken, and the 

procedures to be used, in those cases in 

which individuals who are subject to 

the FFD program: 

(i) Have been involved in the use, 

sale, or possession of illegal substances, 

illegal drugs, or illicit substances; 

(ii) Are impaired by any illegal 

substances, illegal drugs, or illicit 

substances or the consumption of 

alcohol as determined by behavioral 

observation or a test that measures 

blood alcohol concentration; 

(iii) If drug and alcohol testing is 

conducted, attempted to subvert the 

testing process by adulterating or 

diluting specimens (in vivo or in vitro), 

substituting specimens, or by any other 

means; 

(iv) If drug and alcohol testing is 

conducted, refused to provide a 

specimen for analysis or follow 

instructions provided by FFD program 

personnel; 

(v) Had legal action taken relating to 

drug or alcohol use; or 

(vi) Demonstrated character or actions 

indicating that the individual cannot be 

trusted or relied upon to perform those 

duties and responsibilities or maintain 

access to NRC-licensed facilities, special 

nuclear material (SNM), or sensitive 

information. 

(3) The process, including the duties 

and responsibilities of FFD program 

personnel, to be followed if an 

individual’s behavior or condition raises 

a concern regarding the possible use, 

sale, or possession of illegal drugs onor offsite; the possible use or possession 

of alcohol on the NRC-licensed facility; 

impairment from any cause that in any 

way could adversely affect the 

individual’s ability to safely and 

competently perform the individual’s 

duties; or the receipt of credible 

information indicating that the 

individual cannot be trusted or relied on 

to perform those duties and 

responsibilities making the individual 

subject to this part. 

(4) Operation and oversight of an 

onsite or offsite collection facility. 

(5) The fatigue management 

requirements in §§ 26.202(b) and either 

26.205(d)(3) or (d)(7). 

(6) Measures to prevent subversion of 

drug and alcohol tests conducted onsite 

and offsite. 

§ 26.607 Drug and alcohol testing. 

Licensees and other entities 

implementing § 26.604, at their 

discretion, and licensees and other 

entities implementing § 26.605 must 

perform drug and alcohol testing that 

complies with the following 

requirements— 

(a) Split specimens. Split specimen 

collections of oral fluid or urine must be 

used for the test conditions described in 

paragraph (b) of this section. A split 

specimen collection need not be used if 

the licensee or other entity elects to use 

a POCTA device for a screening test 

conducted during random testing under 

paragraphs (b)(2) and (h) of this section 

or a protected area portal monitor 

indication that drugs or alcohol were 

detected under paragraph (j) of this 

section. Testing of the split specimen 

(specimen B) requires the donor’s 

permission unless ordered by the MRO 

to resolve an invalid test result obtained 

for specimen A. 

(b) Test conditions. Individuals 

identified in § 26.4 must be subject to 

drug and alcohol testing under the 

following conditions: 

(1) Pre-access. A pre-access test must 

be conducted for drugs and alcohol 

before performing or directing the 

conduct of roles and responsibilities 

making the individual subject to this 

subpart or being granted unescorted 

access to the protected areas of the NRClicensed facility. A pre-access test must 

have been conducted no more than 14 

days before the individual is granted 

unescorted access. 

(2) Random. Random testing for drugs 

and alcohol must— 

(i) Be administered in a manner that 

provides reasonable assurance that 

individuals are unable to predict the 

time periods during which specimens 

will be collected; 

(ii) Require individuals who are 

selected for random testing to report to 

the onsite collection site as soon as 

reasonably practicable after notification, 

within the time period specified in the 

FFD program procedure; 

(iii) Ensure that all individuals in the 

population that is subject to random 

testing on a given day have an equal 

probability of being selected and tested; 

(iv) Ensure that an individual 

completing a test is immediately eligible 

for another random test; and 

(v) Ensure that the sampling process 

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testing provides that the number of 

random tests performed annually is 

equal to at least 50 percent for licensee 

employees and 50 percent for 

contractor/vendors at the NRC-licensed 

site. 

(3) For-cause. A for-cause drug test, 

alcohol test, or both, must be conducted 

onsite in response to an individual’s 

observed behavior or physical condition 

indicating possible substance abuse or 

after receiving credible information that 

an individual is engaging in substance 

abuse, as defined in § 26.5; 

(4) Post-event. A post-event test for 

drugs and alcohol must be conducted— 

(i) As soon as practical after an event 

involving a human error that was 

committed by an individual specified in 

§ 26.4, where the human error may have 

caused or contributed to the event. This 

test must be conducted onsite unless the 

individual requires offsite medical care. 

The licensee or other entity must test 

the individual(s) who committed or 

directed the error and need not test 

individuals who were affected by the 

event and whose actions likely did not 

cause or contribute to the event. The 

licensee or other entity must describe in 

its procedures what constitutes a human 

error. 

(ii) Within 4 hours of an event unless 

immediate medical intervention 

precludes the conduct of the test on the 

individual(s) who caused or contributed 

to the accident(s), if the event results 

in— 

(A) An illness or personal injury to 

any individual which results in death, 

days away from work, restricted work, 

transfer to another job, medical 

treatment beyond first aid, loss of 

consciousness, or other significant 

illness or injury, as diagnosed by a 

licensee- or other entity-designated 

physician or other licensed health care 

professional, even if the illness or injury 

does not result in death, days away from 

work, restricted work or job transfer, 

medical treatment beyond first aid, or 

loss of consciousness; or 

(B) Damage to any safety- or securityrelated structures, systems, and 

components; and 

(5) Follow-up. An individual subject 

to part 26 who has violated the FFD 

policy for substance use or abuse, or the 

sale, use, or possession of illegal drugs 

must be subject to a follow-up series of 

tests for drugs, alcohol, or both to verify 

an individual’s continued abstinence 

from substance abuse. 

(c) Urine and oral fluid specimens. (1) 

All urine or oral fluid specimens must 

be subject to validity testing, including 

an adulterant and biological marker, and 

tested for the substances listed in 

§ 26.31(d)(1), except as allowed by 

§ 26.603(e)(3). 

(2) For the use of urine as the 

biological specimen to be tested, the 

following requirements must be 

implemented— 

(i) § 26.115, Collecting a urine 

specimen under direct observation; 

(ii) § 26.119, Determining ‘‘shy’’ 

bladder; and 

(iii) § 26.163, Cutoff levels for drugs 

and drug metabolites, (a)(2) regarding 

special analysis testing. 

(3) For alcohol testing onsite, the 

following requirements must be 

implemented— 

(i) § 26.91, Acceptable devices for 

conducting initial and confirmatory 

tests for alcohol and methods of use; 

(ii) § 26.93, Preparing for alcohol 

testing; 

(iii) § 26.95, Conducting an initial test 

for alcohol using a breath specimen; 

(iv) § 26.97, Collecting oral fluid 

specimens for alcohol and drug testing; 

(v) § 26.99, Determining the need for 

a confirmatory test for alcohol; 

(vi) § 26.101, Conducting a 

confirmatory test for alcohol; and, 

(vii) § 26.103, Determining a 

confirmed positive test result for 

alcohol. 

(4) For all test conditions in paragraph 

(b) of this section, except for the use of 

a POCTA screening device in paragraph 

(h) of this section, and for MRO-directed 

tests under § 26.185, drug testing must 

be performed at an HHS-certified 

laboratory for the specific biological 

specimen to be tested. Only HHScertified laboratory test results from 

urine and oral fluid specimens may be 

used for the issuance of a part 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 § 26.603(e), that 

satisfies the requirements of the U.S. 

Food and Drug Administration (FDA) 

for commercial distribution. Specimens 

that yield positive, adulterated, 

substituted, or invalid initial validity or 

drug test results or discrepant biological 

markers must be subject to confirmatory 

testing by an HHS-certified laboratory, 

certified for that biological specimen, 

except for invalid specimens that cannot 

be tested. 

(g) Oral fluid testing. If the licensee or 

other entity elects to use oral fluid for 

drug or alcohol testing, the collection, 

packaging, and temporary storage of the 

drug or alcohol test device, and 

shipment of an oral fluid specimen to an 

HHS-certified laboratory or the 

collection of an oral fluid specimen for 

alcohol testing must be performed in 

accordance with licensee- or other 

entity-established procedures based 

either on the requirements in part 26 or 

the procedures in HHS Guidelines 

identified by the licensee or other entity 

in § 26.606(b)(1)(iv). The device must 

have received premarket approval from 

the FDA and must not expire before 

laboratory testing. The drugs, drug 

metabolites, initial and confirmatory 

testing cutoffs, and biological markers, if 

applicable, must be those established by 

HHS for oral fluid testing and the 

alcohol cutoffs in this part or, if not 

established by HHS or the NRC for the 

panel of drugs and drug metabolites to 

be tested, as determined and 

documented by a forensic toxicologist 

review conducted pursuant to 

§ 26.31(d)(1)(i)(D). 

(h) Point of collection testing and 

assessment. (1) If the licensee or other 

entity elects to use a POCTA device, 

then it may only be used for pre-access 

and random drug and alcohol initial 

testing in paragraph (b) of this section, 

the alcohol testing process in paragraph 

(c)(3) of this section, and the portal area 

screening process in paragraph (j) of this 

section. Before the licensee or other 

entity uses a POCTA device, a forensic 

toxicologist must review and document 


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their evaluation that the validity and 

accuracy of the device for alcohol and/ 

or the drugs and drug metabolites listed 

in § 26.31(d) are comparable to the 

performance achieved by initial testing 

conducted using a similar technology at 

an HHS-certified laboratory. For initial 

testing of drugs and drug metabolites 

using a POCTA device, this review must 

include a documented evaluation of 

POCTA device performance against the 

requirements in § 26.161(b) for a urine 

specimen or the procedures in the HHS 

Guidelines for urine or oral fluid, as 

implemented by the licensee or other 

entity through its procedures. 

(2) If the performance of the POCTA 

device is not comparable to that 

achieved from initial testing conducted 

by an HHS-certified laboratory as 

determined by the forensic toxicologist, 

then the licensee or other entity must 

implement a mitigating strategy to 

maintain program effectiveness under 

§ 26.603(e)(2), as applicable. 

(3) The licensee and other entity must 

implement procedures for the use of a 

POCTA that ensures the effectiveness of 

the collection process, assessment of the 

screening results, and prevention of 

subversion attempts. 

(4) If the use of a POCTA device 

indicates a discrepant biological marker 

or that a test result exceeds the initial 

test cutoff, the specimen is invalid, or 

the individual subverted the drug or 

alcohol test, then the individual must be 

immediately removed from duties, 

responsibilities, and access making the 

individual subject to this subpart. 

(i) The individual must be subject to 

an immediate drug and alcohol test 

using the alcohol testing process in 

paragraph (c)(3) of this section for a 

positive alcohol screen and either oral 

fluid or urine by a collection kit that is 

not a POCTA device, but of the same 

type of biological specimen collected by 

the POCTA, for validity, if required, and 

initial and confirmatory testing by an 

HHS-certified laboratory. 

(ii) If this individual shows any signs 

of impairment, the individual’s 

authorization must be temporarily 

removed until the MRO reviews the 

laboratory test result(s), interviews the 

individual, and performs a 

determination of fitness under § 26.189 

or § 26.619, as applicable, that enables 

the restoration of authorization. 

(i) Hair testing. The testing of hair 

specimens may only be used to inform 

a licensee’s or other entity’s 

determination of whether the individual 

is trustworthy and reliable under the 

test condition in paragraph (b)(1) of this 

section to supplement the information 

gained from a pre-access test using oral 

fluid or urine as the test specimen and 

must be conducted at an HHS-certified 

laboratory certified for hair specimens. 

(1) If used, this process must be 

described in the licensee’s or other 

entity’s FFD policy and described in 

detail in its procedure. The panel of 

drugs and drug metabolites to be 

evaluated must only include those listed 

as Schedule I or II of section 202 of the 

Controlled Substances Act [21 U.S.C. 

812]. The collection, packaging, and 

temporary storage of a hair specimen 

and shipment of the specimen to an 

HHS-certified laboratory must be 

conducted in accordance with the HHS 

Guidelines. The test kit must be FDA 

cleared, and licensee- or other entitydesignated FFD program personnel must 

conduct the collection, packaging, 

temporary storage, shipping, and 

custody and control of the specimen. 

(2) Before the licensee or other entity 

begins to conduct hair testing, the initial 

and confirmatory testing cutoffs must be 

the cutoffs established by HHS for hair 

testing or, if not established by HHS or 

the NRC, as determined by a forensic 

toxicologist review conducted pursuant 

to § 26.31(d)(1)(i)(D). 

(3) Confirmed positive test results 

must be considered potentially 

disqualifying FFD information until 

proven otherwise by a review under 

§ 26.613. Sanctions under this subpart 

must not be issued for any FFD policy 

violation involving a drug test using a 

hair specimen unless the licensee or 

other entity determines that the 

individual subverted, as defined in 

§ 26.5, the hair test. 

(j) Portal area screening. A noninvasive point of collection testing 

instrument may be used to screen 

individuals for drugs, drug metabolites, 

and alcohol before the individuals’ 

entry into or exit from a protected or 

vital area. 

(1) If a licensee or other entity uses 

such an instrument, then before such 

use, a forensic toxicologist must review 

the instrument and document an 

evaluation that the instrument and 

setpoints used in the instrument are 

acceptable for use for the detection and 

screening of the drugs and drug 

metabolites selected for screening from 

the panel of drugs and drug metabolites 

to be tested under the FFD program and 

alcohol and its metabolites. 

(2) The instrument must be operated 

in accordance with the manufacturer’s 

specifications. If screening detects the 

presence of drugs, drug metabolites, or 

alcohol at or above the instrument set 

point(s), the individual screened by the 

instrument must be subject to a POCTA 

screening test using the process 

described in paragraph (h) of this 

section or an oral fluid or urine test that 

is sent to an HHS-certified laboratory. 

(3) A part 26 sanction may not be 

issued to an individual based solely on 

a portal area screening instrument 

detection that drugs or alcohol exceed 

the instrument’s established setpoint. 

(k) Blood testing. The testing of blood 

specimens may only be conducted 

under the order of the licensee- or other 

entity-designated MRO for a valid 

medical reason as confirmed by the 

MRO pursuant to § 26.31(d)(5). This 

specimen must be subject to testing by 

a laboratory that satisfies quality control 

requirements that are comparable to 

those required for certification by the 

HHS. 

(l) Custody-and-control form. For the 

collection and packaging of urine, oral 

fluid, and hair specimens, the licensee 

or other entity must use a custody-andcontrol form approved by the U.S. 

Office of Management and Budget. For 

the use of a POCTA device, the licensee 

or other entity must implement a 

licensee- or other entity-approved and 

-maintained procedure that ensures the 

reliability of the tracking, handling, and 

storage of a specimen from the point of 

specimen collection to the final 

disposition of the specimen and the 

reliability of an identification system to 

uniquely assign the specimen to the 

donor. 

(m) Medical Review Officer. Licensees 

or other entities must— 

(1) Require their designated MRO to 

review positive, adulterated, 

substituted, and dilute confirmatory 

drug and validity test results and test 

results of questionable validity to 

determine whether the donor has 

violated the FFD policy for urine and 

oral fluid specimens. The review must 

be completed before reporting the 

results to the individual designated by 

the licensee or other entity to assess 

authorization or perform the suitability 

and fitness determinations required 

under § 26.619, or, if required, that are 

described in subpart H of this part. 

(2) Require their MRO to satisfy the 

requirements in § 26.183 and, prior to 

conducting any activities under this 

part, attend and pass a medical- or 

clinical-based training session to 

improve his/her knowledge of MRO 

duties and responsibilities, drug and 

alcohol testing processes and 

procedures, and evaluation of drug 

testing results. This training session 

must be conducted by a nationally 

recognized MRO training and 

certification organization that has been 

assessed by the licensee’s or other 

entity’s FFD program personnel to 

include the technical elements an MRO 

must implement under § 26.185. An 


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MRO who performed the duties and 

responsibilities in §§ 26.185 and 26.187 

for at least 3 continuous years in the last 

10 years prior to being hired or 

contracted by the licensee or other 

entity satisfies the requirements in this 

paragraph. 

(3) Require their MRO to attend a 

medical- or clinical-based training 

session on a triennial basis to improve 

his/her knowledge of changes in drug 

and alcohol testing processes and 

procedures and evaluation of drug 

testing results. 

(4) Require their MRO to determine 

whether a biological specimen is 

positive, adulterated, substituted, dilute 

or of questionable validity by 

implementing the requirements in 

§ 26.185 or the HHS Guidelines through 

the licensee’s or other entity’s 

procedures. 

(i) If § 26.185 or the HHS Guidelines, 

as used by the licensee or other entity 

in its procedures, are insufficient to 

make this determination, then guidance 

issued by a State agency in the state in 

which the NRC-licensed facility is 

located, Federal agencies, or nationally 

recognized MRO training and 

certification organizations may be used 

to inform an MRO determination. 

(ii) An MRO need not review a 

confirmed alcohol positive test result 

determined by an evidentiary breath 

testing device under paragraphs 

(c)(3)(vi) and (vii) of this section. 

(5) Require their MRO to determine 

and approve the use of oral fluid or 

urine as an alternative biological 

specimen when the donor cannot 

provide a specimen for testing. This 

determination and the retest must be 

documented and completed as soon as 

reasonably practicable. 

(6) Require the MRO to review all 

specimens screened and tested 

associated with a drug-related FFD 

policy violation. This review includes 

POCTA, split specimens, and all 

specimens taken to resolve a discrepant 

condition, such as a possible subversion 

attempt, impairment without a known 

cause, or a donor-requested or MROdirected re-test. To resolve a discrepant 

condition, the MRO is authorized to test 

a specimen for a biological marker, 

adulterants, or additional drugs. 

(n) Limitations of screening and 

testing. Specimens collected under NRC 

regulations may only be designated or 

approved for screening and testing as 

described in this part and may not be 

used to conduct any other analysis or 

test without the written permission of 

the donor. Analyses, screens, and tests 

that may not be conducted include, but 

are not limited to, DNA testing, 

serological typing, or any other medical 

or genetic test used for diagnostic or 

specimen identification purposes. No 

biological specimens may be passively 

sampled and analyzed in a manner 

different than described in this subpart. 

(o) Specimen collectors. All onsite 

specimen collections, except a 

collection by a portal area screening 

instrument in paragraph (j) of this 

section, must be conducted by licenseeor other entity-designated and -trained 

personnel. 

§ 26.608 FFD program training. 

(a) FFD program training. (1) 

Individuals must be trained in the FFD 

policy and procedure, including fatigue 

management, and their FFD program 

responsibilities. Individuals who collect 

specimens for testing or screening must 

also be trained in specimen collector 

duties and responsibilities, including, at 

a minimum, specimen collection, 

custody and control, identification and 

response to subversion attempts, and 

privacy. For licensees and other entities 

of commercial nuclear plants, the FFD 

program training program must use a 

systems approach to training as defined 

in § 53.725 of this chapter and described 

in § 53.830 of this chapter for those 

individuals in § 26.4. 

(2) FFD program training must 

include training on the behavioral 

observation program. The behavioral 

observation program training must 

include the detection of physiological 

behaviors or conditions that may 

indicate— 

(i) Possible use, sale, or possession of 

illegal drugs or illicit drugs, or 

substance abuse on- or offsite; 

(ii) Use or possession of alcohol onsite 

or use while on duty offsite; 

(iii) Impairment from fatigue or any 

cause that, if left unattended, could 

result in inattentiveness or human 

errors; and 

(iv) Any individual’s inability to 

safely and competently perform 

assigned duties and responsibilities or 

act in a trustworthy and reliable manner 

while having access to protected areas, 

SNM, or sensitive information. 

(3) Training must explain that an 

individual’s FFD policy violation will— 

(i) Subject the individual to an FFD 

program-required sanction designed to 

preclude recurrence of an FFD policy 

violation; 

(ii) Contribute to the licensee’s or 

other entity’s assessment of whether the 

individual can be trusted and relied 

upon to safely and competently perform 

the assigned duties and responsibilities 

making the individual subject to this 

subpart; 

(iii) Be used to inform the licensee’s 

or other entity’s insider mitigation and 

access authorization programs under 

§§ 73.55, 73.56, 73.100 or 73.120 of this 

chapter; and 

(iv) Be used to inform other NRC 

licensees and other entities subject to 

part 26 when FFD program information 

is requested to support authorization 

determinations under subpart C of this 

part or §§ 73.56 or 73.120 of this 

chapter. 

(b) Training and assessments. 

Training and a trainee assessment must 

be conducted before pre-access testing, 

and refresher training and trainee 

assessments must be conducted 

periodically thereafter. 

(c) Training program review. The 

licensee or other entity must 

periodically evaluate its FFD training 

program and revise it as appropriate to 

reflect industry experience as well as 

applicable changes to the regulations in 

this part, the HHS Guidelines, if used, 

and specimen collection and testing 

processes implemented by the licensee 

or other entity. 

§ 26.609 Behavioral observation. 

(a) Licensees and other entities must 

ensure that the individuals who are 

subject to this subpart are subject to 

behavioral observation and that 

behavioral observation is performed by 

all individuals subject to this subpart. 

(b) Licensees and other entities must 

require all individuals subject to the 

FFD program to report to the licenseeor other entity-designated representative 

any onsite or offsite behaviors or 

activities by individuals subject to this 

part that may constitute an 

unreasonable risk to the safety or 

security of the NRC-licensed facility or 

SNM or may cause harm to others. This 

reporting must include any information 

relating to character or reputation of the 

individual indicating that the individual 

cannot be trusted or relied upon to 

perform those duties and 

responsibilities or maintain access to 

NRC-licensed facilities, SNM, or 

sensitive information that makes them 

subject to part 26. 

(c) Behavioral observation must be 

performed visually, in-person, and, 

when necessary, remotely by live video 

and audible streaming and capture, to 

observe the behavior of individuals in 

the workforce subject to the 

requirements in this subpart. 

(d) Not withstanding paragraph (c) of 

this section, for a reactor facility where 

individual task loading does not allow 

for the effective conduct of behavior 

observation in addition to assigned 

operational tasks, the licensee or other 

entity must implement a live video and 

audible streaming and capture system to 

conduct behavioral observation of 


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persons licensed to operate under 10 

CFR part 53 who manipulate the 

controls of any commercial nuclear 

plant licensed under 10 CFR part 53. 

§ 26.610 Sanctions. 

Licensees and other entities that 

implement an FFD program under this 

subpart must establish sanctions for 

FFD policy violations that, at a 

minimum, prohibit the individuals 

specified in § 26.4 from being assigned 

to perform or direct those duties and 

responsibilities or maintaining 

authorization making them subject to 

this subpart. The severity of the 

sanction must escalate with the number 

of occurrences and severity of the FFD 

policy violation. The sanction must be 

long enough to act as a deterrent and, 

if the individual is retained as a licensee 

employee or contractor/vendor, 

facilitate the individual to complete 

counseling or treatment. The sanctions 

must include a minimum 5-year denial 

of access to the NRC-licensed facility for 

any individual who is determined to 

have been involved in the sale, use, or 

possession of illegal drugs or the 

consumption of alcohol within a 

protected area of any facility licensed 

under part 53 of this chapter or within 

a transporter’s facility or vehicle used in 

the conveyance of formula quantities of 

strategic SNM while the individual is 

subject to this subpart, and a permanent 

denial of access to the NRC-licensed 

facility for three FFD policy violations 

or any subversion attempt of any drug 

or alcohol test or screening process, 

including subversion attempts at any 

licensee or other entity subject to this 

part. 

§ 26.611 Protection of information. 

(a) Licensees and other entities that 

collect personal information about an 

individual for the purpose of complying 

with this subpart must establish and 

maintain a system of files and 

procedures to prevent unauthorized 

disclosure. 

(b) Licensees and other entities must 

obtain a signed consent that documents 

the individual’s acceptance of being 

subject to the FFD program and 

authorizes the disclosure of the personal 

information collected and maintained 

under this subpart, except for 

disclosures to the individuals and 

entities specified in § 26.37(b)(1) 

through (b)(6), (b)(8), and persons 

deciding matters under review in 

§ 26.613. This signed and dated consent 

must be obtained before making the 

individual subject to the FFD program. 

§ 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 

§ 26.4 has violated the FFD policy. The 

procedure must provide for an objective 

and impartial review of the facts related 

to the determination that the individual 

has violated the FFD policy and a 

schedule for the completion of the 

review. 

§ 26.615 Audits. 

(a) Licensees and other entities that 

implement an FFD program under this 

subpart must audit their programs at a 

frequency that ensures the continuing 

effectiveness of their FFD program, FFD 

program elements that are provided by 

C/Vs, and the FFD programs of C/Vs 

that are accepted by the licensee or 

other entity. Corrective actions must be 

as soon as reasonably practicable to 

resolve any problems identified in an 

audit and preclude recurrence. 

(b) The subject matter, scope, and 

frequency of audits must be revised as 

necessary to improve or maintain 

program performance based on findings 

resulting from licensee or other entity 

implementation of its FFD PMRP in 

§ 26.603(d). 

(c) Licensees and other entities may 

conduct joint audits or accept audits of 

C/Vs so long as the audit addresses the 

relevant services of the C/Vs. 

(d) Licensees and other entities must 

audit HHS-certified laboratories unless 

the licensee’s or other entity’s panel of 

drugs and drug metabolites to be tested 

is equivalent to the panel by which the 

laboratory is certified by HHS or is 

subject to the standards and procedures 

for drug testing and evaluation used by 

the laboratory under the HHS 

Guidelines. Licensees and other entities 

must audit any hospital or other facility 

licensed by the State (or Statedesignated entity) if used to conduct 

specimen collections and perform 

alcohol testing under this part on a 

biennial basis to confirm that the facility 

procedures are comparable to those 

described in subpart E of this part, for 

urine and oral fluid. 

§ 26.617 Recordkeeping and reporting. 

(a) Licensees and other entities that 

implement FFD programs under this 

subpart must ensure that records 

pertaining to the administration of their 

program, which may be stored and 

archived electronically, are maintained 

so that they are available for NRC 

inspection purposes and for any legal 

proceedings resulting from the 

administration of the program. Records 

pertaining to the administration of the 

FFD program and FFD performance data 

required by § 26.717 must be retained 

until license termination. 

(b) Licensees and other entities must 

make the following reports: 

(1) Reports to the NRC Operations 

Center by telephone within 24 hours 

after the licensee or other entity 

discovers any intentional act that casts 

doubt on the integrity of the FFD 

program and any programmatic failure, 

degradation, or discovered vulnerability 

of the FFD program that may permit 

undetected drug or alcohol use or abuse 

by individuals who are subject to this 

subpart. These events must be reported 

under this subpart, rather than under 

the provisions of § 73.1200 of this 

chapter; and 

(2) Annual program performance 

reports for the FFD program, including 

the FFD program performance data 

listed in § 26.717(b), as applicable. 

Licensees and other entities must 

submit FFD program performance data 

(for January through December) to the 

NRC annually, before March 1 of the 

following year and must use unexpired 

NRC-provided forms for the electronic 

submission of FFD information to the 

NRC. 

(c) Licensees and other entities 

subject to this subpart must describe in 

sufficient detail to support an 

authorization determination, an 

individual’s FFD policy violation (while 

protecting privacy information under 

§ 26.611) and FFD program weakness to 

NRC, licensees, and other entities 

subject to this part when requested to 

support authorization determinations 

under subpart C of this part or § 73.120 

of this chapter, as applicable, or to 

support licensee or other entity 

performance monitoring. 

§ 26.619 Suitability and fitness 

determinations. 

Licensees and other entities that 

implement FFD programs under this 

subpart must develop, implement, and 

maintain procedures for evaluating 

whether to assign individuals to 

perform or direct those duties and 

responsibilities making them subject to 

this subpart. A suitability or fitness 

determination conducted for cause must 

be performed face to face. A suitability 

or fitness determination conducted for 

cause may be performed remotely using 

electronic communications only when 

supported by someone who is present 

in-person with the individual being 

assessed, and that supporting person 

must be trained in accordance with the 

requirements of either §§ 26.29 or 

26.608. 

■ 94. Revise § 26.709 to read as follows: 


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§ 26.709 Applicability. 

(a) The requirements of this subpart 

apply to the FFD programs of licensees 

and other entities specified in § 26.3(a) 

through (d), except for FFD programs 

that are implemented under subpart K 

of this part. 

(b) The requirements in this subpart 

apply to the FFD programs of licensees 

and other entities specified in § 26.3(f) 

that elect not to implement the 

requirements in subpart M or elect to 

implement the requirements in 

§ 26.605(b). 

§ 26.711 [Amended] 

■ 95. In § 26.711, in paragraphs (c) and 

(d), remove the phrase ‘‘(c) and (d),’’ and 

add in its place the phrase ‘‘(c), (d), and 

(f),’’. 

§ 26.825 [Amended] 

■ 96. In § 26.825, in paragraph (b) add 

remove the phrase ‘‘§§ 26.1, 26.3, 26.5, 

26.7, 26.8, 26.9, 26.11, 26.51, 26.81, 

26.121, 26.151, 26.181, 26.201, 26.823, 

and 26.825’’ and add in its place the 

phrase ‘‘§§ 26.1, 26.3, 26.5, 26.7, 26.8, 

26.9, 26.11, 26.51, 26.81, 26.121, 26.151, 

26.181, 26.201, 26.601, 26.823, and 

26.825’’. 

PART 30—RULES OF GENERAL 

APPLICABILITY TO DOMESTIC 

LICENSING OF BYPRODUCT 

MATERIAL 

■ 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. 

■ 98. In § 30.4, revise the definition for 

‘‘Utilization facility’’ to read as follows: 

§ 30.4 Definitions. 

* * * * * 

Utilization facility means a utilization 

facility as defined in the regulations 

contained in part 50 or part 53 of this 

chapter; 

■ 99. In § 30.50, revise paragraph (c)(3) 

to read as follows: 

§ 30.50 Reporting requirements. 

* * * * * 

(c) * * * 

(3) The provisions of this section do 

not apply to licensees subject to the 

notification requirements in §§ 50.72 or 

53.1630 of this chapter. They do apply 

to those part 50 licensees possessing 

material licensed under this part, who 

are not subject to the notification 

requirements in § 50.72 of this chapter. 

PART 40—DOMESTIC LICENSING OF 

SOURCE MATERIAL 

■ 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. 

■ 101. In § 40.60, revise paragraph (c)(3) 

to read as follows: 

§ 40.60 Reporting requirements. 

* * * * * 

(c) * * * 

(3) The provisions of this section do 

not apply to licensees subject to the 

notification requirements in §§ 50.72 or 

53.1630 of this chapter. They do apply 

to those part 50 licensees possessing 

material licensed under this part who 

are not subject to the notification 

requirements in § 50.72 of this chapter. 

PART 50—DOMESTIC LICENSING OF 

PRODUCTION AND UTILIZATION 

FACILITIES 

■ 102. The authority citation for part 50 

continues to read as follows: 

Authority: Atomic Energy Act of 1954, 

secs. 11, 101, 102, 103, 104, 105, 108, 122, 

147, 149, 161, 181, 182, 183, 184, 185, 186, 

187, 189, 223, 234 (42 U.S.C. 2014, 2131, 

2132, 2133, 2134, 2135, 2138, 2152, 2167, 

2169, 2201, 2231, 2232, 2233, 2234, 2235, 

2236, 2237, 2239, 2273, 2282); Energy 

Reorganization Act of 1974, secs. 201, 202, 

206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); 

Nuclear Waste Policy Act of 1982, sec. 

306(42 U.S.C. 10226); National 

Environmental Policy Act of 1969 (42 U.S.C. 

4332); 44 U.S.C. 3504 note; Sec. 109, Pub. L. 

96–295, 94 Stat. 783. 

■ 103. In § 50.47, revise paragraphs 

(a)(1) and (e) to read as follows: 

§ 50.47 Emergency plans. 

(a)(1)(i) Except as provided in 

paragraph (d) of this section, no initial 

operating license for a nuclear power 

reactor will be issued under this part or 

under part 53 of this chapter unless a 

finding is made by the NRC that there 

is reasonable assurance that adequate 

protective measures can and will be 

taken in the event of a radiological 

emergency. No finding under this 

section is necessary for issuance of a 

renewed nuclear power reactor 

operating license. 

(ii) No initial combined license under 

parts 52 or 53 of this chapter will be 

issued unless a finding is made by the 

NRC that there is reasonable assurance 

that adequate protective measures can 

and will be taken in the event of a 

radiological emergency. No finding 

under this section is necessary for 

issuance of a renewed combined 

license. 

(iii) If an application for an early site 

permit under subpart A of part 52 of this 

chapter includes complete and 

integrated emergency plans under 

§ 52.17(b)(2)(ii) of this chapter or an 

application for an early site permit 

under subpart H of part 53 of this 

chapter includes complete and 

integrated emergency plans under 

§ 53.1146(b)(2)(ii) of this chapter, no 

early site permit will be issued unless 

a finding is made by the NRC that the 

emergency plans provide reasonable 

assurance that adequate protective 

measures can and will be taken in the 

event of a radiological emergency. 

(iv) If an application for an early site 

permit proposes major features of the 

emergency plans under §§ 52.17(b)(2)(i) 

or 53.1146(b)(2)(i) of this chapter, no 

early site permit will be issued unless 

a finding is made by the NRC that the 

major features are acceptable in 

accordance with the applicable 

standards of either § 50.47 and appendix 

E to this part or the applicable 

requirements of § 50.160, within the 

scope of emergency preparedness 

matters addressed in the major features. 

* * * * * 

(e) Notwithstanding the requirements 

of paragraph (b) of this section and the 

provisions of § 52.103 or § 53.1452 of 

this chapter, a holder of a combined 

license under part 52 or part 53 of this 

chapter, as applicable, that is complying 

with the requirements of § 50.47(b) and 

appendix E to this part may not load 

fuel or operate except as provided in 

accordance with appendix E to this part 

and § 50.54(gg), and a holder of a 

combined license under part 52 or part 

53 of this chapter that is complying with 

the requirements of § 50.160 may not 

load fuel or operate except as provided 

in accordance with § 50.160(c)(2) and 

§ 50.54(gg). 

* * * * * 

■ 104. In § 50.54, revise paragraphs 

(q)(2), (q)(4), and (gg)(1) introductory 

text to read as follows: 

§ 50.54 Conditions of licenses. 

* * * * * 

(q) * * * 

(2)(i) Except as provided in paragraph 

(q)(2)(ii) of this section, a holder of a 

license under this part, or a combined 

license under parts 52 or 53 of this 

chapter after the Commission makes the 

finding under §§ 52.103(g) or 53.1452(g) 


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of this chapter, as applicable, shall 

follow and maintain the effectiveness of 

an emergency plan that meets the 

requirements in appendix E to this part 

and, for nuclear power reactor licensees, 

the planning standards of § 50.47(b). 

(ii) A holder of a license under this 

part for a non-power production or 

utilization facility, a holder of a license 

under this part or part 53 of this chapter 

for a small modular reactor or a nonlight-water reactor, or a holder of a 

combined license under parts 52 or 53 

of this chapter after the Commission 

makes the finding under §§ 52.103(g) or 

53.1452(g) of this chapter, as applicable, 

for a small modular reactor or a nonlight-water reactor, shall follow and 

maintain the effectiveness of either an 

emergency plan that meets the 

requirements in § 50.160 or an 

emergency plan that meets the 

requirements in appendix E to this part 

and, for nuclear power reactor licensees, 

the planning standards of § 50.47(b). 

* * * * * 

(4) The changes to a licensee’s 

emergency plan that reduce the 

effectiveness of the plan as defined in 

paragraph (q)(1)(iv) of this section may 

not be implemented without prior 

approval by the NRC. A licensee 

desiring to make such a change shall 

submit an application for an 

amendment to its license. In addition to 

the filing requirements of §§ 50.90 and 

50.91 or §§ 53.1510 and 53.1515 of this 

chapter, as applicable, the request must 

include all emergency plan pages 

affected by that change and must be 

accompanied by a forwarding letter 

identifying the change, the reason for 

the change, and the basis for concluding 

that the licensee’s emergency plan, as 

revised, will continue to meet either the 

requirements in § 50.160 or the 

requirements in appendix E to this part 

and, for nuclear power reactor licensees, 

the planning standards of § 50.47(b). 

* * * * * 

(gg)(1) Notwithstanding §§ 52.103 or 

53.1452 of this chapter, if following the 

conduct of the exercise required by 

paragraph IV.f.2.a of appendix E to this 

part or § 50.160(c)(2), as applicable, 

FEMA identifies one or more 

deficiencies in the state of offsite 

emergency preparedness, the holder of a 

combined license under 10 CFR part 52 

or under 10 CFR part 53, as applicable, 

may operate at up to 5 percent of rated 

thermal power only if the Commission 

finds that the state of onsite emergency 

preparedness provides reasonable 

assurance that adequate protective 

measures can and will be taken in the 

event of a radiological emergency. The 

NRC will base this finding on its 

assessment of the applicant’s onsite 

emergency plans against the pertinent 

standards in either § 50.47 and 

appendix E to this part, or § 50.160, as 

applicable. Review of the applicant’s 

emergency plans will include the 

following standards with offsite aspects: 

* * * * * 

■ 105. In § 50.160, revise paragraphs 

(b)(3) and (c)(2) to read as follows: 

§ 50.160 Emergency preparedness for 

small modular reactors, non-light-water 

reactors, and non-power production or 

utilization facilities. 

* * * * * 

(b) * * * 

(3) Emergency planning zone. For an 

applicant whose analysis required by 

§ 50.33(g)(2) or § 53.1109(g)(2) of this 

chapter meets the criteria in 

§ 50.33(g)(2)(i) or § 53.1109(g)(2)(i) of 

this chapter, as applicable, determine 

and describe the boundary and physical 

characteristics of the EPZ in the 

emergency plan. 

* * * * * 

(c) * * * 

(2) A holder of a combined license 

issued under parts 52 or 53 of this 

chapter before the Commission has 

made the finding under §§ 52.103(g) or 

53.1452(g) of this chapter, as applicable, 

must establish, implement, and 

maintain an emergency preparedness 

program that meets the requirements of 

paragraph (b) of this section, as 

described in the approved emergency 

plan and license, and conduct an initial 

exercise to demonstrate this compliance 

within 2 years before the scheduled date 

for initial loading of fuel (or, for a fueled 

manufactured reactor, within 2 years 

before the scheduled date for initiating 

the physical removal of any one of the 

independent physical mechanisms to 

prevent criticality required under 

§ 53.620(d)(1) of this chapter). 

■ 106. In appendix B to part 50, revise 

the first paragraph in the Introduction 

section, the first paragraph of section III, 

Design Control, and section IV, 

Procurement Document Control, to read 

as follows: 

Appendix B to Part 50—Quality 

Assurance Criteria for Nuclear Power 

Plants and Fuel Reprocessing Plants 

Introduction. Every applicant for a 

construction permit is required by the 

provisions of § 50.34 or § 53.1309 of this 

chapter to include in its preliminary safety 

analysis report a description of the quality 

assurance program to be applied to the 

design, fabrication, construction, and testing 

of the structures, systems, and components of 

the facility. Every applicant for an operating 

license is required by the provisions of 

§ 50.34 or § 53.1369 of this chapter to 

include, in its final safety analysis report, 

information pertaining to the managerial and 

administrative controls to be used to assure 

safe operation. Every applicant for a 

combined license is required by the 

provisions of §§ 52.79 or 53.1416 of this 

chapter to include in its final safety analysis 

report a description of the quality assurance 

applied to the design, and to be applied to 

the fabrication, construction, and testing of 

the structures, systems, and components of 

the facility and to the managerial and 

administrative controls to be used to assure 

safe operation. For applications submitted 

after September 27, 2007, every applicant for 

an early site permit is required by the 

provisions of §§ 52.17 or 53.1146 of this 

chapter to include in its site safety analysis 

report a description of the quality assurance 

program applied to site activities related to 

the design, fabrication, construction, and 

testing of the structures, systems, and 

components of a facility or facilities that may 

be constructed on the site. Every applicant 

for a design approval is required by the 

provisions of §§ 52.137 or 53.1209 of this 

chapter to include in its final safety analysis 

report a description of the quality assurance 

program applied to the design of the 

structures, systems, and components of the 

facility. Every applicant for a design 

certification is required by the provisions of 

§§ 52.47 or 53.1239 of this chapter to include 

in its final safety analysis report a description 

of the quality assurance program applied to 

the design of the structures, systems, and 

components of the facility. Every applicant 

for a manufacturing license is required by the 

provisions of §§ 52.157 or 53.1279 of this 

chapter to include in its final safety analysis 

report a description of the quality assurance 

program applied to the design, and to be 

applied to the manufacture of, the structures, 

systems, and components of the reactor. 

Nuclear power plants and fuel reprocessing 

plants include structures, systems, and 

components that prevent or mitigate the 

consequences of postulated accidents that 

could cause undue risk to the health and 

safety of the public. This appendix 

establishes quality assurance requirements 

for the design, manufacture, construction, 

and operation of those structures, systems, 

and components. The pertinent requirements 

of this appendix apply to all activities 

affecting the safety-related functions of those 

structures, systems, and components; these 

activities include designing, purchasing, 

fabricating, handling, shipping, storing, 

cleaning, erecting, installing, inspecting, 

testing, operating, maintaining, repairing, 

refueling, and modifying. 

* * * * * 

III. Design Control 

Measures shall be established to assure that 

applicable regulatory requirements and the 

design bases, as defined in § 50.2 and as 

specified in the license application, or the 

functional design criteria, as defined in 

§ 53.020 of this chapter and as specified in 

the license application, for those structures, 

systems, and components to which this 

appendix applies are correctly translated into 

specifications, drawings, procedures, and 

instructions. These measures shall include 

provisions to assure that appropriate quality 


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standards are specified and included in 

design documents and that deviations from 

such standards are controlled. Measures shall 

also be established for the selection and 

review for suitability of application of 

materials, parts, equipment, and processes 

that are essential to the safety-related 

functions of the structures, systems and 

components. 

* * * * * 

IV. Procurement Document Control 

Measures shall be established to assure that 

applicable regulatory requirements, design 

bases or functional design criteria, and other 

requirements which are necessary to assure 

adequate quality are suitably included or 

referenced in the documents for procurement 

of material, equipment, and services, whether 

purchased by the applicant or by its 

contractors or subcontractors. To the extent 

necessary, procurement documents shall 

require contractors or subcontractors to 

provide a quality assurance program 

consistent with the pertinent provisions of 

this appendix. 

* * * * * 

PART 51—ENVIRONMENTAL 

PROTECTION REGULATIONS FOR 

DOMESTIC LICENSING AND RELATED 

REGULATORY FUNCTIONS 

■ 107. The authority citation for part 51 

continues to read as follows: 

Authority: Atomic Energy Act of 1954, 

secs. 161, 193 (42 U.S.C. 2201, 2243); Energy 

Reorganization Act of 1974, secs. 201, 202 

(42 U.S.C. 5841, 5842); National 

Environmental Policy Act of 1969 (42 U.S.C. 

4332, 4334, 4335); Nuclear Waste Policy Act 

of 1982, secs. 144(f), 121, 135, 141, 148 (42 

U.S.C. 10134(f), 10141, 10155, 10161, 10168); 

44 U.S.C. 3504 note. 

■ 108. In § 51.20, revise paragraphs 

(b)(1) and (2) to read as follows: 

§ 51.20 Criteria for and identification of 

licensing and regulatory actions requiring 

environmental impact statements. 

* * * * * 

(b) * * * 

(1) Issuance of a limited work 

authorization or a permit to construct a 

nuclear power reactor, testing facility, or 

fuel reprocessing plant under part 50 of 

this chapter, issuance of an early site 

permit under part 52 of this chapter, or 

issuance of a limited work 

authorization, construction permit, or 

early site permit under part 53 of this 

chapter. 

(2) Issuance or renewal of a full power 

or design capacity license to operate a 

nuclear power reactor, testing facility, or 

fuel reprocessing plant under parts 50 or 

53 of this chapter, or a combined license 

under parts 52 or 53 of this chapter. 

* * * * * 

■ 109. In § 51.22, revise paragraphs 

(c)(3) introductory text, (c)(9) 

introductory text, (c)(12) introductory 

text, (c)(17), (c)(22) and (23) to read as 

follows: 

§ 51.22 Criterion for categorical exclusion; 

identification of licensing and regulatory 

actions eligible for categorical exclusion or 

otherwise not requiring environmental 

review. 

* * * * * 

(c) * * * 

(3) Amendments to parts 20, 30, 31, 

32, 33, 34, 35, 37, 39, 40, 50, 51, 52, 53, 

54, 60, 61, 63, 70, 71, 72, 73, 74, 81, and 

100 of this chapter which relate to— 

* * * * * 

(9) Issuance of an amendment to a 

permit or license for a reactor under part 

50, part 52, or part 53 of this chapter 

that changes a requirement or issuance 

of an exemption from a requirement, 

with respect to installation or use of a 

facility component located within the 

restricted area, as defined in part 20 of 

this chapter; or the issuance of an 

amendment to a permit or license for a 

reactor under part 50, part 52, or part 53 

of this chapter that changes an 

inspection or a surveillance 

requirement; provided that: 

* * * * * 

(12) Issuance of an amendment to a 

license under parts 50, 52, 53, 60, 61, 

63, 70, 72, or 75 of this chapter relating 

solely to safeguards matters (i.e., 

protection against sabotage or loss or 

diversion of special nuclear material) or 

issuance of an approval of a safeguards 

plan submitted under parts 50, 52, 53, 

70, 72, and 73 of this chapter, provided 

that the amendment or approval does 

not involve any significant construction 

impacts. These amendments and 

approvals are confined to— 

* * * * * 

(17) Issuance of an amendment to a 

permit or license under part 30, part 40, 

part 50, part 52, part 53, or part 70 of 

this chapter which deletes any limiting 

condition of operation or monitoring 

requirement based on or applicable to 

any matter subject to the provisions of 

the Federal Water Pollution Control Act. 

* * * * * 

(22) Issuance of a standard design 

approval under part 52 or part 53 of this 

chapter. 

(23) The Commission finding for a 

combined license under § 52.103(g) or 

§ 53.1452(g) of this chapter. 

* * * * * 

§ 51.26 [Amended] 

■ 110. In § 51.26, in paragraph (d) 

remove the phrase ‘‘under part 52’’ and 

add in its place the phrase ‘‘under 10 

CFR parts 52 or 53,’’. 

■ 111. In § 51.30, revise paragraph (a) 

introductory text and paragraphs (d) and 

(e) to read as follows: 

§ 51.30 Environmental assessment. 

(a) An environmental assessment for 

proposed actions, other than those for a 

standard design certification under 10 

CFR parts 52 or 53, or a manufacturing 

license under 10 CFR parts 52 or 53, 

shall identify the proposed action and 

include: 

* * * * * 

(d) An environmental assessment for 

a standard design certification under 

subpart B of part 52 of this chapter, or 

under subpart H of part 53 of this 

chapter must identify the proposed 

action and will be limited to the 

consideration of the costs and benefits 

of severe accident mitigation design 

alternatives and the bases for not 

incorporating severe accident mitigation 

design alternatives in the design 

certification. An environmental 

assessment for an amendment to a 

design certification will be limited to 

the consideration of whether the design 

change which is the subject of the 

proposed amendment renders a severe 

accident mitigation design alternative 

previously rejected in the earlier 

environmental assessment to become 

cost beneficial, or results in the 

identification of new severe accident 

mitigation design alternatives, in which 

case the costs and benefits of new severe 

accident mitigation design alternatives 

and the bases for not incorporating new 

severe accident mitigation design 

alternatives in the design certification 

must be addressed. 

(e) An environmental assessment for a 

manufacturing license under subpart F 

of part 52 of this chapter or under 

subpart H of part 53 of this chapter must 

identify the proposed action and will be 

limited to the consideration of the costs 

and benefits of severe accident 

mitigation design alternatives and the 

bases for not incorporating severe 

accident mitigation design alternatives 

in the manufacturing license. An 

environmental assessment for an 

amendment to a manufacturing license 

will be limited to consideration of 

whether the design change which is the 

subject of the proposed amendment 

either renders a severe accident 

mitigation design alternative previously 

rejected in an environmental assessment 

to become cost beneficial, or results in 

the identification of new severe accident 

mitigation design alternatives, in which 

case the costs and benefits of new severe 

accident mitigation design alternatives 

and the bases for not incorporating new 

severe accident mitigation design 

alternatives in the manufacturing 

license must be addressed. In either 

case, the environmental assessment will 

not address the environmental impacts 


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associated with manufacturing the 

reactor under the manufacturing license. 

§ 51.31 [Amended] 

■ 112. In § 51.31, in paragraph (a) 

remove the phrase ‘‘under part 52’’ and 

add in its place the phrase ‘‘under parts 

52 or 53’’. 

§ 51.32 [Amended] 

■ 113. In § 51.32, in paragraphs (b)(1) 

and (3) remove the phrase ‘‘of part 52 of 

this chapter’’ and add in its place the 

phrase ‘‘of part 52 of this chapter or 

subpart H of part 53 of this chapter’’. 

§ 51.49 [Amended] 

■ 114. In § 51.49, in paragraph (c) 

introductory text, remove the phrase ‘‘of 

part 52 of this chapter’’ and add in its 

place the phrase ‘‘of part 52 of this 

chapter or under subpart H of part 53 of 

this chapter’’. 

§ 51.50 [Amended] 

■ 115. In § 51.50, wherever it appears, 

remove the phrase ‘‘in accordance with 

§ 50.36b of this chapter’’ and add in its 

place the phrase ‘‘in accordance with 

§§ 50.36b or 53.1112 of this chapter’’. 

§ 51.53 [Amended] 

■ 116. In § 51.53, in paragraph (d) 

remove the phrase ‘‘under § 50.82 of this 

chapter’’ and add in its place the phrase 

‘‘under §§ 50.82 or 53.1080 of this 

chapter’’. 

§ 51.54 [Amended] 

■ 117. In § 51.54, in paragraph (a), 

remove the phrase ‘‘of part 52 of this 

chapter’’ and add in its place the phrase 

‘‘of part 52 of this chapter or under 

subpart H of part 53 of this chapter’’. 

§ 51.55 [Amended] 

■ 118. In § 51.55, in paragraph (a) 

remove the phrase ‘‘of part 52 of this 

chapter’’ and add in its place the phrase 

‘‘of part 52 of this chapter or under 

subpart H of part 53 of this chapter’’. 

■ 119. In § 51.58, revise paragraph (b) to 

read as follows: 

§ 51.58 Environmental report—number of 

copies; distribution. 

* * * * * 

(b) Each applicant for a license to 

manufacture a nuclear power reactor, or 

for an amendment to a license to 

manufacture, seeking approval of the 

final design of the nuclear power reactor 

under subpart F of part 52 of this 

chapter or under subpart H of part 53 of 

this chapter, shall submit to the 

Commission an environmental report or 

any supplement to an environmental 

report in the manner specified in §§ 52.3 

or 53.040 of this chapter. The applicant 

shall maintain the capability to generate 

additional copies of the environmental 

report or any supplement to the 

environmental report for subsequent 

distribution to parties and Boards in the 

NRC proceeding; Federal, State, and 

local officials; and any affected Indian 

Tribes, in accordance with written 

instructions issued by the Director, 

Office of Nuclear Reactor Regulation. 

■ 120. In § 51.77, revise paragraph (a) 

introductory text to read as follows: 

§ 51.77 Distribution of draft environmental 

impact statement. 

(a) In addition to the distribution 

authorized by § 51.74, a copy of a draft 

environmental statement for a licensing 

action for a production or utilization 

facility, except an action authorizing 

issuance, amendment, or renewal of a 

license to manufacture a nuclear power 

reactor pursuant to 10 CFR part 52, 

subpart F or 10 CFR part 53, subparts H 

or I will also be distributed to: 

* * * * * 

§ 51.92 [Amended] 

■ 121. In § 51.92, in paragraph (b), 

wherever it may appear, remove the 

phrase ‘‘10 CFR part 52’’ and add in its 

place the phrase ‘‘10 CFR parts 52 or 

53’’. 

§ 51.95 [Amended] 

■ 122. In § 51.95, in paragraph (c) 

introductory text remove the phrase 

‘‘under 10 CFR parts 52 or 54’’ and add 

in its place the phrase ‘‘under 10 CFR 

parts 52, 53, or 54’’. 

■ 123. In § 51.101, revise paragraph 

(a)(2) to read as follows: 

§ 51.101 Limitations on actions. 

(a) * * * 

(2) Any action concerning the 

proposal taken by an applicant which 

would— 

(i) Have an adverse environmental 

impact, or 

(ii) Limit the choice of reasonable 

alternatives that may be grounds for 

denial of the license. In the case of an 

application covered by §§ 30.32(f), 

40.31(f), 50.10(c), 53.1130, 70.21(f), or 

72.16 and 72.34 of this chapter, the 

provisions of this paragraph will be 

applied in accordance with 

§ 30.33(a)(5), 40.32(e), 50.10(c), 53.1130, 

70.23(a)(7), or 72.40(b) of this chapter, 

as appropriate. 

* * * * * 

§ 51.103 [Amended] 

■ 124. In § 51.103, in paragraph (a)(6) 

remove the phrase ‘‘under 10 CFR 

50.10’’ and add in its place the phrase 

‘‘under §§ 50.10 or 53.1130 of this 

chapter’’. 

■ 125. In § 51.105, revise paragraph 

(c)(1) introductory text to read as 

follows: 

§ 51.105 Public hearings in proceedings 

for issuance of construction permits or 

early site permits; limited work 

authorizations. 

* * * * * 

(c)(1) In addition to complying with 

the applicable provisions of § 51.104, in 

any proceeding for the issuance of a 

construction permit for a nuclear power 

plant or an early site permit under parts 

52 or 53 of this chapter, where the 

applicant requests a limited work 

authorization under §§ 50.10(d) or 

53.1130 of this chapter, the presiding 

officer will— 

* * * * * 

■ 126. In § 51.107, revise paragraphs (a) 

introductory text, (b) introductory text, 

and (d)(1) introductory text to read as 

follows: 

§ 51.107 Public hearings in proceedings 

for issuance of combined licenses; limited 

work authorizations. 

(a) In addition to complying with the 

applicable requirements of § 51.104, in 

a proceeding for the issuance of a 

combined license for a nuclear power 

reactor under parts 52 or 53 of this 

chapter, the presiding officer will: 

* * * * * 

(b) If a combined license application 

references an early site permit, then the 

presiding officer in the combined 

license hearing must not admit any 

contention proffered by any party on 

environmental issues that have been 

accorded finality under §§ 52.39 or 

53.1188 of this chapter, unless the 

contention: 

* * * * * 

(d)(1) In any proceeding for the 

issuance of a combined license where 

the applicant requests a limited work 

authorization under §§ 50.10(d) or 

§ 53.1130(a) of this chapter, the 

presiding officer, in addition to 

complying with any applicable 

provision of § 51.104, will: 

* * * * * 

■ 127. Revise § 51.108 to read as 

follows: 

§ 51.108 Public hearings on Commission 

findings that inspections, tests, analyses, 

and acceptance criteria of combined 

licenses are met. 

In any public hearing requested under 

§§ 52.103(b) or 53.1452(b) of this 

chapter, the Commission will not admit 

any contentions on environmental 

issues, the adequacy of the 

environmental impact statement for the 

combined license issued under subpart 

C of part 52 of this chapter or under 


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subpart H of part 53 of this chapter, or 

the adequacy of any other 

environmental impact statement or 

environmental assessment referenced in 

the combined license application. The 

Commission will not make any 

environmental findings in connection 

with the finding under § 52.103(g) or 

§ 53.1452(g) of this chapter. 

■ 128. Add part 53, consisting of 

§§ 53.000 through 53.9010, to read as 

follows: 

PART 53—RISK-INFORMED, 

TECHNOLOGY-INCLUSIVE 

REGULATORY FRAMEWORK FOR 

COMMERCIAL NUCLEAR PLANTS 

Sec. 

53.000 Purpose. 

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 designbasis accidents. 

53.425 Design features and functional 

design criteria for normal operations. 

53.430 Design features and functional 

design criteria for protection of plant 

workers. 

53.440 Design requirements. 

53.450 Analysis requirements. 

53.460 Safety categorization and treatments. 

53.470 Maintaining analytical safety 

margins used to justify operational 

flexibilities. 

53.480 Earthquake engineering. 

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-reliantmitigation facilities. 

53.805 Facility licensee requirements 

related to generally licensed reactor 

operators. 

53.810 Generally licensed reactor operators. 

53.815 Generally licensed reactor operator 

training, examination, and proficiency 

programs. 

53.820 Cessation of individual 

applicability. 

53.830 Training and qualification of 

commercial nuclear plant personnel. 

53.845 Programs. 

53.850 Radiation protection. 

53.855 Emergency preparedness. 

53.860 Security programs. 

53.865 Quality assurance. 

53.870 Integrity assessment programs. 

53.875 Fire protection. 

53.880 Inservice inspection and inservice 

testing. 

53.910 Procedures and guidelines. 

Subpart G—Decommissioning 

Requirements 

53.1000 Scope and purpose. 

53.1010 Financial assurance for 

decommissioning. 

53.1020 Cost estimates for 

decommissioning. 

53.1030 Annual adjustments to cost 

estimates for decommissioning. 

53.1040 Methods for providing financial 

assurance for decommissioning. 

53.1045 Limitations on the use of 

decommissioning trust funds. 

53.1050 NRC oversight. 

53.1060 Reporting and recordkeeping 

requirements. 

53.1070 Termination of license. 

53.1075 Program requirements during 

decommissioning. 

53.1080 Release of part of a commercial 

nuclear plant or site for unrestricted use. 

Subpart H—Licenses, Certifications, and 

Approvals 

53.1100 Filing of application for licenses, 

certifications, or approvals; oath or 

affirmation. 

53.1101 Requirement for license. 

53.1103 Combining applications and 

licenses. 

53.1106 Elimination of repetition. 

53.1109 Contents of applications; general 

information. 

53.1112 Environmental conditions. 

53.1115 Agreement limiting access to 

classified information. 

53.1118 Ineligibility of certain applicants. 

53.1120 Exceptions and exemptions from 

licensing requirements. 

53.1121 Public inspection of applications. 

53.1124 Relationship between sections. 

53.1130 Limited work authorizations. 

53.1140 Early site permits. 

53.1143 Filing of applications. 

53.1144 Contents of applications for early 

site permits; general information. 

53.1146 Contents of applications for early 

site permits; technical information. 

53.1149 Review of applications. 

53.1155 Referral to the Advisory Committee 

on Reactor Safeguards. 

53.1158 Issuance of early site permit. 

53.1161 Extent of activities permitted. 

53.1164 Duration of permit. 

53.1167 Limited work authorization after 

issuance of early site permit. 

53.1170 Transfer of early site permit. 

53.1173 Application for renewal. 

53.1176 Criteria for renewal. 

53.1179 Duration of renewal. 

53.1182 Use of site for other purposes. 

53.1188 Finality of early site permit 

determinations. 

53.1200 Standard design approvals. 

53.1203 Filing of applications. 

53.1206 Contents of applications for 

standard design approvals; general 

information. 

53.1209 Contents of applications for 

standard design approvals; technical 

information. 

53.1210 Contents of applications for 

standard design approvals; other 

application content. 

53.1212 Standards for review of 

applications. 


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53.1215 Referral to the Advisory Committee 

on Reactor Safeguards. 

53.1218 Staff approval of design. 

53.1221 Finality of standard design 

approvals; information requests. 

53.1230 Standard design certifications. 

53.1233 Filing of applications. 

53.1236 Contents of applications for 

standard design certifications; general 

information. 

53.1239 Contents of applications for 

standard design certifications; technical 

information. 

53.1241 Contents of applications for 

standard design certifications; other 

application content. 

53.1242 Review of applications. 

53.1245 Referral to the Advisory Committee 

on Reactor Safeguards. 

53.1248 Issuance of standard design 

certification. 

53.1251 Duration of certification. 

53.1254 Application for renewal. 

53.1257 Criteria for renewal. 

53.1260 Duration of renewal. 

53.1263 Finality of standard design 

certifications. 

53.1270 Manufacturing licenses. 

53.1273 Filing of applications. 

53.1276 Contents of applications for 

manufacturing licenses; general 

information. 

53.1279 Contents of applications for 

manufacturing licenses; technical 

information. 

53.1282 Contents of applications for 

manufacturing licenses; other 

application content. 

53.1285 Review of applications. 

53.1286 Referral to the Advisory Committee 

on Reactor Safeguards. 

53.1287 Issuance of manufacturing licenses. 

53.1288 Finality of manufacturing licenses. 

53.1291 Duration of manufacturing 

licenses. 

53.1293 Transfer of manufacturing licenses. 

53.1295 Renewal of manufacturing licenses. 

53.1300 Construction permits. 

53.1306 Contents of applications for 

construction permits; general 

information. 

53.1309 Contents of applications for 

construction permits; technical 

information. 

53.1312 Contents of applications for 

construction permits; other application 

content. 

53.1315 Review of applications. 

53.1318 Finality of referenced NRC 

approvals, permits, and certifications. 

53.1324 Referral to the Advisory Committee 

on Reactor Safeguards. 

53.1327 Authorization to conduct limited 

work authorization activities. 

53.1330 Exemptions, departures, and 

variances. 

53.1333 Issuance of construction permits. 

53.1336 Finality of construction permits. 

53.1342 Duration of construction permits. 

53.1345 Transfer of construction permits. 

53.1348 Termination of construction 

permits. 

53.1360 Operating licenses. 

53.1366 Contents of applications for 

operating licenses; general information. 

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. 

§ 53.000 Purpose. 

This part provides an optional 

technology-inclusive, performancebased framework for the issuance, 

amendment, renewal, and termination 

of licenses, permits, certifications, and 

approvals for commercial nuclear plants 

licensed under section 103 of the 

Atomic Energy Act of 1954, as amended 

(the Act)(68 Stat. 919), and Title II of the 

Energy Reorganization Act of 1974, as 

amended (88 Stat. 1242). Also, this part 

gives notice to all persons who 

knowingly provide to any holder of or 

applicant for an approval, certification, 

permit, or license, or to a contractor, 

subcontractor, or consultant of any of 


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them, components, equipment, 

materials, or other goods or services that 

relate to the activities of a holder of or 

applicant for an approval, certification, 

permit, or license, subject to this part, 

that they may be individually subject to 

U.S. Nuclear Regulatory Commission 

enforcement action for violation of the 

provisions in § 53.050. 

Subpart A—General Provisions 

§ 53.015 Scope. 

Subpart A provides general provisions 

applicable to all applicants and 

licensees subject to the rules of this part. 

§ 53.020 Definitions. 

For the purpose of this part: 

Anticipated event sequence means 

event sequences expected to occur one 

or more times during the life of a 

commercial nuclear plant. Anticipated 

event sequences take into account the 

expected response of all structures, 

systems, and components (SSCs) within 

the plant, regardless of safety 

classification. 

Applicant means a person applying 

for a license, permit, or other form of 

Commission permission or approval 

under this part. 

Certified fuel handler means, for a 

commercial nuclear plant, either— 

(1) A non-licensed operator who has 

qualified in accordance with a fuel 

handler training program approved by 

the Commission; or 

(2) A non-licensed operator who 

demonstrates compliance with the 

following criteria: 

(i) Has qualified in accordance with a 

fuel handler training program that 

demonstrates compliance with the same 

requirements as training programs for 

non-licensed operators required by 

§ 53.830, and 

(ii) Is responsible for decisions on— 

(A) Safe conduct of decommissioning 

activities, 

(B) Safe handling and storage of spent 

fuel, and 

(C) Appropriate response to plant 

emergencies. 

Combined license (COL) means a 

combined construction permit (CP) and 

operating license (OL) with conditions 

for a commercial nuclear plant issued 

under this part. 

Commercial nuclear plant means a 

facility consisting of one or more 

commercial nuclear reactors and 

associated co-located support facilities, 

including the collection of buildings, 

radionuclide sources, and SSCs for 

which a license, certification, or 

approval is being sought under this part, 

that is or will be used for producing 

power for commercial electric power or 

other commercial purposes. For the 

purposes of requirements in this part 

that reference requirements in part 50 of 

this chapter, a commercial nuclear plant 

is equivalent to a nuclear power plant. 

Commercial nuclear reactor means an 

apparatus, other than an atomic 

weapon, designed or used to sustain 

nuclear fission. For the purposes of 

requirements in this part that reference 

requirements in 10 CFR part 50, a 

commercial nuclear reactor is 

equivalent to a nuclear reactor as 

defined in 10 CFR 50.2. 

Commission means the U.S. Nuclear 

Regulatory Commission (NRC) or its 

duly authorized representatives. 

Consensus code or standard means 

any technical standard that is— 

(1) Developed or adopted by a 

voluntary consensus standard body 

under procedures that assure that 

persons having interests within the 

scope of the standard that are affected 

by the provisions of the standard have 

reached substantial agreement on its 

adoption; 

(2) Formulated in a manner that 

afforded an opportunity for diverse 

views to be considered; and 

(3) Designated by the standards body 

as a consensus code or standard. 

Construction means the activities in 

paragraph (1) below and does not mean 

the activities in paragraph (2) below. 

(1) Activities constituting 

construction are those activities credited 

or relied upon for demonstrating 

compliance with the safety criteria 

defined in subpart B of this part which 

are conducted on-site to build the 

commercial nuclear plant, including the 

driving of piles; subsurface preparation; 

placement of backfill, concrete, or 

permanent retaining walls within an 

excavation; installation of foundations; 

or in-place assembly, erection, 

fabrication, or testing, which are for— 

(i) Safety-related (SR) and non-safetyrelated but safety-significant (NSRSS) 

SSCs of a facility; 

(ii) SSCs necessary to comply with 10 

CFR part 73; or 

(iii) Onsite emergency facilities 

necessary to comply with § 53.855. 

(2) Construction does not include— 

(i) Changes for temporary use of the 

land for public recreational purposes; 

(ii) Site exploration, including 

necessary borings to determine 

foundation conditions or other 

preconstruction monitoring to establish 

background information related to the 

suitability of the site, the environmental 

impacts of construction or operation, or 

the protection of environmental values; 

(iii) Preparation of a site for 

construction of a facility, including 

clearing of the site, grading, installation 

of drainage, erosion, and other 

environmental mitigation measures, and 

construction of temporary roads and 

borrow areas; 

(iv) Erection of fences and other 

access control measures; 

(v) Excavation; 

(vi) Erection of support buildings 

(such as construction equipment storage 

sheds, warehouse and shop facilities, 

utilities, concrete mixing plants, 

docking and unloading facilities, and 

office buildings) for use in connection 

with the construction of the facility; 

(vii) Building of service facilities 

(such as paved roads, parking lots, 

railroad spurs, exterior utility and 

lighting systems, potable water systems, 

sanitary sewage treatment facilities, and 

transmission lines); 

(viii) Procurement or fabrication of 

components or portions of the proposed 

facility occurring at locations other than 

the final, in-place location at the 

facility; or 

(ix) Manufacture of a nuclear power 

reactor under a manufacturing license 

(ML) under subpart H of this part to be 

installed at the proposed site and to be 

part of the proposed facility. 

Custom combined license (custom 

COL) means a COL that does not 

reference a standard design approval or 

design certification. 

Decommission or decommissioning 

means to remove a plant or site safely 

from service and reduce residual 

radioactivity to a level that permits— 

(1) Release of the property for 

unrestricted use and termination of the 

license; or 

(2) Release of the property under 

restricted conditions and termination of 

the license. 

Defense in depth means inclusion of 

two or more independent and 

redundant layers of defense in the 

design of a facility and its operating 

procedures to compensate for 

uncertainties such that no single layer of 

defense, no matter how robust, is 

exclusively relied upon. Defense in 

depth includes, but is not limited to, the 

use of access controls, physical barriers, 

redundant and diverse safety functions, 

and emergency response measures. 

Design-basis accidents (DBAs) means 

postulated event sequences that are 

used to set functional design criteria 

and performance objectives for the 

design of SR SSCs through deterministic 

analyses. Design-basis accidents are a 

type of licensing-basis event and are 

based on the capabilities and 

reliabilities of SR SSCs needed to 

mitigate and prevent event sequences, 

respectively. 

Design-basis external hazard level 

means the level of severity or intensity 


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of an external hazard for which the SR 

SSCs are protected against or designed 

to withstand without losing their 

capability to perform their safety 

functions. 

Design features means the active and 

passive SSCs and the inherent 

characteristics of those SSCs that 

contribute to limiting the total effective 

dose equivalent to individual members 

of the public during normal operations 

and prevent or mitigate the 

consequences of event sequences. 

Electric utility means any entity that 

generates or distributes electricity and 

that recovers the cost of this electricity, 

either directly or indirectly, through 

rates established by the entity itself or 

by a separate regulatory authority. 

Investor-owned utilities, including 

generation or distribution subsidiaries, 

public utility districts, municipalities, 

rural electric cooperatives, and State 

and Federal agencies, including 

associations of any of the foregoing, are 

included within the meaning of 

‘‘electric utility.’’ 

Event sequence means a postulated 

initiating event defined for a set of 

initial plant conditions followed by 

system, safety function, and operator 

successes or failures, and terminating in 

a specified end state depending on the 

system, safety function, and operator 

successes and failures (e.g., prevention 

of release of radioactive material or 

release in one of the reactor-specific 

release categories). An event sequence 

may include many unique variations of 

events that are similar in terms of 

results or end states. 

Exclusion area means that area 

surrounding the reactor, in which the 

reactor licensee has the authority to 

determine all activities including 

exclusion or removal of personnel and 

property from the area. This area may be 

traversed by a highway, railroad, or 

waterway, provided these are not so 

close to the facility as to interfere with 

normal operations of the facility and 

provided appropriate and effective 

arrangements are made to control traffic 

on the highway, railroad, or waterway, 

in case of emergency, to protect the 

public health and safety. Residence 

within the exclusion area must normally 

be prohibited. In any event, residents 

must be subject to ready removal in case 

of necessity. Activities unrelated to 

operation of the reactor may be 

permitted in an exclusion area under 

appropriate limitations, provided that 

no significant hazards to the public 

health and safety will result. 

Fission product release means the 

amount and composition of radioactive 

material released to the environment, 

after accounting for any retention of 

radionuclides provided by reactor 

design features. 

Fuel means special nuclear material 

(SNM) or source material, discrete 

elements that physically contain SNM 

or source material, and homogeneous 

mixtures that contain SNM or source 

material, intended to or used to create 

power in a commercial nuclear plant. 

Functional design criteria means 

metrics for the performance of SSCs. For 

SR SSCs, these criteria define 

performance metrics necessary to 

demonstrate compliance with the safety 

criteria in § 53.210. For NSRSS SSCs, 

these criteria define performance 

metrics necessary to demonstrate 

compliance with the safety criteria in 

§ 53.220. 

License, when used in the context of 

a facility, means a limited work 

authorization, CP, OL, early site permit, 

COL, or ML under this part, or a 

renewed license issued by the 

Commission under this part. When used 

in the context of a license authorizing 

an individual to manipulate the controls 

of a facility, license means a license 

issued by the Commission to perform 

the function of an operator, senior 

operator, or generally licensed reactor 

operator as defined in this part. 

Licensee means a person who is 

authorized to conduct activities under a 

license issued under this part by the 

Commission. 

Licensing-basis events means a 

collection of event sequences 

considered in the design and licensing 

of the commercial nuclear plant. 

Licensing-basis events are unplanned 

events and include anticipated event 

sequences, unlikely event sequences, 

very unlikely event sequences, and 

DBAs. 

Licensing-basis information means the 

information contained in regulations, 

orders, licenses, certifications, or 

approvals issued by the NRC for a 

commercial nuclear plant licensed 

under this part and that information 

submitted to the NRC by an applicant or 

licensee in a Safety Analysis Report, 

program description, or other licensingrelated document required under this 

part. 

Low-population zone means the area 

immediately surrounding the exclusion 

area which contains residents, the total 

number and density of which are such 

that there is a reasonable probability 

that appropriate protective measures 

could be taken on their behalf in the 

event of a serious accident. A 

permissible population density or total 

population within this zone is not 

included in this definition because the 

situation may vary from case to case. 

Whether a specific number of people 

can, for example, be evacuated from a 

specific area or instructed to take shelter 

on a timely basis, will depend on many 

factors such as location, number and 

size of highways, scope and extent of 

advance planning, and actual 

distribution of residents within the area. 

Major decommissioning activity 

means, for a commercial nuclear plant, 

any activity that results in permanent 

removal of major radioactive 

components, permanently modifies the 

structure of the containment, if 

applicable, or results in dismantling 

components for shipment containing 

greater than class C waste in accordance 

with 10 CFR 61.55. 

Major feature of the emergency plans 

means an aspect of those plans 

necessary to: 

(1) Address in whole or part either 

one or more of the 16 standards in 10 

CFR 50.47(b) or the requirements of 10 

CFR 50.160(b), as applicable; or 

(2) Describe the emergency planning 

zones as required in § 53.1109(g). 

Manufactured reactor means the 

essential portions of a nuclear reactor 

that are manufactured under an ML and 

subsequently transported and 

incorporated into a commercial nuclear 

plant under a COL. 

Manufacturing license means a 

license issued under this part that 

authorizes the manufacture of 

manufactured reactors but not its 

construction, installation, or operation. 

Non-Safety-Related but SafetySignificant (NSRSS) SSCs means those 

SSCs which are not SR but are relied on 

to achieve adequate defense in depth or 

perform risk-significant functions and 

warrant special treatment. 

Non-Safety-Significant SSCs means 

those SSCs that are not SR or NSRSS, 

are not relied on to achieve adequate 

defense in depth or to perform risksignificant functions, and do not 

warrant special treatment. 

Person means— 

(1) any individual, corporation, 

partnership, firm, association, trust, 

estate, public or private institution, 

group, government agency other than 

the Commission or the Department, 

except that the Department shall be 

considered a person to the extent that its 

facilities are subject to the licensing and 

related regulatory authority of the 

Commission pursuant to section 202 of 

the ERA, any State or any political 

subdivision of, or any political entity 

within a State, any foreign government 

or nation or any political subdivision of 

any such government or nation, or other 

entity; and 

(2) any legal successor, representative, 

agent, or agency of the foregoing. 


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Population center distance means the 

distance from the reactor to the nearest 

boundary of a densely populated center 

containing more than about 25,000 

residents. 

Probabilistic risk assessment means a 

quantitative assessment of the risk 

associated with plant operation and 

maintenance that is measured in terms 

of event sequence occurrence 

frequencies and consequences. 

Programmatic controls means 

administrative procedures that govern 

human action in implementing 

programs and operating, monitoring, 

and maintaining SSCs and equipment of 

a commercial nuclear plant. 

Programmatic controls considered to be 

licensing basis information are specified 

in an application for a requested activity 

of the Commission. 

Quality assurance (QA) means all 

those planned and systematic actions 

necessary to ensure that a structure, 

system, or component will perform 

satisfactorily in service. Quality 

assurance includes quality control, 

which comprises those QA actions 

related to the physical characteristics of 

a material, structure, component, or 

system which provide a means to 

control the quality of the material, 

structure, component, or system to 

predetermined requirements. 

Safety criteria means performancebased metrics that establish a level of 

safety provided in requirements in 

§§ 53.210 and 53.220. 

Safety-related structures, systems, or 

components means those SSCs that are 

relied upon to demonstrate compliance 

with the safety criteria in § 53.210 and 

warrant special treatment. 

Small modular reactor means a power 

reactor, which may be of modular 

design as defined in 10 CFR 52.1, 

licensed under this part to produce heat 

energy up to 1,000 megawatts thermal 

per module. 

Site characteristics means the actual 

physical, environmental, and 

demographic features of a site. Site 

characteristics are specified in an early 

site permit or in a Preliminary or Final 

Safety Analysis Report for a limited 

work authorization, CP, or COL, as 

applicable. 

Site parameters are the postulated 

physical, environmental, and 

demographic features of an assumed 

site. Site parameters are specified in a 

standard design approval, standard 

design certification, or ML. 

Source material means source 

material as defined in subsection 11z. of 

the Atomic Energy Act of 1954, as 

amended, (the Act) and in the 

regulations contained in part 40 of this 

chapter. 

Special nuclear material (SNM) 

means: 

(1) Plutonium, uranium-233, uranium 

enriched in the isotope-233 or in the 

isotope-235, and any other material 

which the Commission, pursuant to the 

provisions of section 51 of the Act, 

determines to be SNM, but does not 

include source material; or 

(2) Any material artificially enriched 

by any of the foregoing, but does not 

include source material. 

Special treatment means those 

requirements, such as QA and 

programmatic controls, that ensure that 

SR and NSRSS SSCs will provide 

defense in depth or perform risksignificant functions. The requirements 

also ensure that the SSCs will perform 

under the service conditions and with 

the reliability assumed in the analysis 

performed under § 53.450 to 

demonstrate compliance with the safety 

criteria in §§ 53.210 and 53.220. 

Standard design means a design 

which is sufficiently detailed and 

complete to support certification or 

approval in accordance with subpart H 

of this part, and which is usable under 

of this part for a multiple number of 

units or at a multiple number of sites 

without reopening or repeating the 

review. 

Standard design approval or design 

approval means an NRC staff approval, 

issued under subpart H of this part, of 

a final standard design for a commercial 

nuclear plant. The approval may be for 

either the final design for the entire 

reactor facility or the final design of 

major portions thereof. 

Standard design certification or 

design certification means a 

Commission approval, issued under 

subpart H of this part, of a final standard 

design for a nuclear power facility. This 

design may be referred to as a certified 

standard design. 

Total effective dose equivalent means 

the sum of the effective dose equivalent 

(for external exposures) and the 

committed effective dose equivalent (for 

internal exposures). 

Utilization facility means any 

commercial nuclear reactor other than 

one designed or used primarily for the 

formation of plutonium or uranium-233. 

Unlikely event sequences means event 

sequences that are not expected to occur 

in the life of a commercial nuclear plant 

and are less likely than anticipated 

event sequences, but are infrequent 

rather than rare. Unlikely event 

sequences take into account the 

expected response of all SSCs within 

the plant regardless of safety 

classification. 

Very unlikely event sequences means 

event sequences that are not expected to 

occur in the life of a commercial nuclear 

plant, are less likely than an unlikely 

event sequence, and are rare. Very 

unlikely event sequences take into 

account the expected response of all 

SSCs within the plant regardless of 

safety classification. 

§ 53.030 [Reserved] 

§ 53.040 Written communications. 

(a) General requirements. All 

correspondence, reports, applications, 

and other written communications from 

the applicant or licensee to the NRC 

concerning the regulations in this part 

or individual license conditions must be 

sent either by mail addressed: ATTN: 

Document Control Desk, U.S. Nuclear 

Regulatory Commission, Washington, 

DC 20555–0001; by hand delivery to the 

NRC’s offices at 11555 Rockville Pike, 

Rockville, Maryland, between the hours 

of 8:15 a.m. and 4 p.m. eastern time; or, 

where practicable, by electronic 

submission, for example, via Electronic 

Information Exchange, email, or CD– 

ROM. Electronic submissions must be 

made in a manner that enables the NRC 

to receive, read, authenticate, distribute, 

and archive the submission, and process 

and retrieve it a single page at a time. 

Detailed guidance on making electronic 

submissions can be obtained by visiting 

the NRC’s website at http://

www.nrc.gov/site-help/esubmittals.html; by email to 

MSHD.Resource@nrc.gov; or by writing 

the Office of the Chief Information 

Officer, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555– 

0001. The guidance discusses, among 

other topics, the formats the NRC can 

accept, the use of electronic signatures, 

and the treatment of nonpublic 

information. If the communication is on 

paper, the signed original must be sent. 

If a submission due date falls on a 

Saturday, Sunday, or Federal holiday, 

the next Federal working day becomes 

the official due date. 

(b) Distribution requirements. Copies 

of all correspondence, reports, and other 

written communications concerning the 

regulations in this part or individual 

license conditions, or the terms and 

conditions of an early site permit or 

standard design approval, must be 

submitted to the persons listed below 

(addresses for the NRC Regional Offices 

are listed in appendix D to 10 CFR part 

20). 

(1) Applications for amendment of 

permits and licenses, reports, and other 

communications. All written 

communications (including responses to 

generic letters, bulletins, information 

notices, regulatory information 

summaries, inspection reports, and 


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miscellaneous requests for additional 

information) that are required of holders 

of licenses, permits, and design 

approvals issued pursuant to this part, 

must be submitted as follows, except as 

otherwise specified in paragraphs (b)(2) 

through (7) of this section: to the NRC’s 

Document Control Desk (if on paper, the 

signed original), with a copy to the 

appropriate Regional Office, and a copy 

to the appropriate NRC Resident 

Inspector if one has been assigned to the 

site of the facility or the place of 

manufacture of a reactor licensed under 

this part. 

(2) Applications for permits and 

licenses, and amendments to 

applications. Applications for licenses, 

permits, and design approvals and 

amendments to any of these types of 

applications must be submitted to the 

NRC’s Document Control Desk, with a 

copy to the appropriate Regional Office, 

and a copy to the appropriate NRC 

Resident Inspector if one has been 

assigned to the facility or the place of 

manufacture of a reactor licensed under 

this part, except as otherwise specified 

in paragraphs (b)(3) through (9) of this 

section. If the application or amendment 

is on paper, the submission to the 

Document Control Desk must be the 

signed original. 

(3) Acceptance review application. 

Written communications required for an 

application for determination of 

suitability for docketing must be 

submitted to the NRC’s Document 

Control Desk, with a copy to the 

appropriate Regional Office. If the 

communication is on paper, the 

submission to the Document Control 

Desk must be the signed original. 

(4) Security plan and related 

submissions. Written communications, 

as defined in paragraphs (b)(4)(i) 

through (v) of this section, must be 

submitted to the NRC’s Document 

Control Desk, with a copy to the 

appropriate Regional Office. If the 

communication is on paper, the 

submission to the Document Control 

Desk must be the signed original. 

Submissions should include the 

following as appropriate: 

(i) Physical security plan; 

(ii) Safeguards contingency plan; 

(iii) Cybersecurity plan; 

(iv) Change to security plan, guard 

training and qualification plan, 

safeguards contingency plan, or 

cybersecurity plan made without prior 

Commission approval under § 53.1565; 

and 

(v) Application for amendment of 

physical security plan, guard training 

and qualification plan, safeguards 

contingency plan, or cybersecurity plan 

under § 53.1510. 

(5) Emergency plan and related 

submissions. Written communications 

as defined in paragraphs (b)(5)(i) 

through (iii) of this section must be 

submitted to the NRC’s Document 

Control Desk, with a copy to the 

appropriate Regional Office, and a copy 

to the appropriate NRC Resident 

Inspector if one has been assigned to the 

site of the facility. If the communication 

is on paper, the submission to the 

Document Control Desk must be the 

signed original. Submissions should 

include the following as appropriate: 

(i) Emergency plan; 

(ii) Change to an emergency plan 

under § 53.1565; and 

(iii) Emergency implementing 

procedures under § 53.855. 

(6) Updated Final Safety Analysis 

Report. An Updated Final Safety 

Analysis Report or replacement pages 

under § 53.1545 must be submitted to 

the NRC’s Document Control Desk, with 

a copy to the appropriate Regional 

Office, and a copy to the appropriate 

NRC Resident Inspector if one has been 

assigned to the site of the facility or the 

place of manufacture of a reactor 

licensed under this part. Paper copy 

submissions may be made using 

replacement pages; however, if a 

licensee chooses to use electronic 

submission, all subsequent updates or 

submissions must be performed 

electronically on a total replacement 

basis. If the communication is on paper, 

the submission to the Document Control 

Desk must be the signed original. If the 

communications are submitted 

electronically, see Guidance for 

Electronic Submissions to the 

Commission. 

(7) Quality assurance related 

submissions. (i) A change to the Safety 

Analysis Report QA program 

description under § 53.1565, or a change 

to a licensee’s NRC-accepted QA topical 

report under § 53.1565, must be 

submitted to the NRC’s Document 

Control Desk, with a copy to the 

appropriate Regional Office, and a copy 

to the appropriate NRC Resident 

Inspector if one has been assigned to the 

site of the facility or the place of 

manufacture of a reactor licensed under 

this part. If the communication is on 

paper, the submission to the Document 

Control Desk must be the signed 

original. 

(ii) A change to an NRC-accepted QA 

topical report from non-licensees (i.e., 

architect/engineers, nuclear steam 

supply system suppliers, fuel suppliers, 

constructors, etc.) must be submitted to 

the NRC’s Document Control Desk. If 

the communication is on paper, the 

signed original must be sent. 

(8) Certification of permanent 

cessation of operations. The licensee’s 

certification of permanent cessation of 

operations, under subpart G of this part, 

must state the date on which operations 

have ceased or will cease, and must be 

submitted to the NRC’s Document 

Control Desk. This submission must be 

under oath or affirmation. 

(9) Certification of permanent fuel 

removal. The licensee’s certification of 

permanent fuel removal, under subpart 

G of this part, must state the date on 

which the fuel was removed from the 

reactor vessel and the disposition of the 

fuel, and must be submitted to the 

NRC’s Document Control Desk. This 

submission must be under oath or 

affirmation. 

(c) Form of communications. All 

paper copies submitted to demonstrate 

compliance with the requirements set 

forth in paragraph (b) of this section 

must be typewritten, printed, or 

otherwise reproduced in permanent 

form on unglazed paper. Exceptions to 

these requirements imposed on paper 

submissions may be granted for the 

submission of micrographic, 

photographic, or similar forms. 

(d) Regulation governing submission. 

Licensees, applicants, and holders of 

standard design approvals submitting 

correspondence, reports, and other 

written communications under the 

regulations of this part are requested but 

not required to cite whenever practical, 

in the upper right corner of the first 

page of the submission, the specific 

regulation or other basis requiring 

submission. 

§ 53.050 Deliberate misconduct. 

(a) Any licensee or applicant for a 

license; holder of or applicant for a 

standard design approval; applicant for 

a standard design certification; 

employee of a licensee, holder of a 

standard design approval, or applicant 

for a license, standard design approval, 

or standard design certification; or any 

contractor (including a supplier or 

consultant), subcontractor, employee of 

a contractor or subcontractor of any 

licensee or applicant for a license, 

holder of or applicant for a standard 

design approval, or applicant for a 

standard design certification, who 

knowingly provides to any licensee, 

applicant, contractor, or subcontractor, 

any components, equipment, materials, 

or other goods or services that relate to 

a licensee’s or applicant’s activities in 

this part, may not— 

(1) Engage in deliberate misconduct 

that causes or would have caused, if not 

detected, a licensee or applicant to be in 

violation of any rule, regulation, or 

order; or any term, condition, or 


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limitation of any license issued by the 

Commission; or 

(2) Deliberately submit to the NRC, a 

licensee, an applicant, or a licensee’s or 

applicant’s contractor or subcontractor, 

information that the person submitting 

the information knows to be incomplete 

or inaccurate in some respect material to 

the NRC. 

(b) A person who violates paragraph 

(a)(1) or (2) of this section may be 

subject to enforcement action in 

accordance with the procedures in 

subpart B of 10 CFR part 2. 

(c) For the purposes of paragraph 

(a)(1) of this section, deliberate 

misconduct by a person means an 

intentional act or omission that the 

person knows— 

(1) Would cause a licensee or 

applicant to be in violation of any rule, 

regulation, or order; or any term, 

condition, or limitation, of any license 

issued by the Commission; or 

(2) Constitutes a violation of a 

requirement, procedure, instruction, 

contract, purchase order, or policy of a 

licensee, applicant, contractor, or 

subcontractor. 

§ 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 § 19.11(e)(1) of this 

chapter. This form must be posted at 

locations sufficient to permit employees 

protected by this section to observe a 

copy on the way to or from their place 

of work. Premises must be posted no 

later than 30 days after an application 

is docketed and remain posted while the 

application is pending before the 

Commission, during the term of the 

license, and for 30 days following 

license termination. 

(2) Copies of NRC Form 3 may be 

obtained by writing to the Regional 

Administrator of the appropriate NRC 

Regional Office listed in appendix D to 

10 CFR part 20, via email to 

Forms.Resource@nrc.gov, or by visiting 

the NRC’s online library at http://

www.nrc.gov/reading-rm/doccollections/forms/. 

(f) No agreement affecting the 

compensation, terms, conditions, or 

privileges of employment, including an 

agreement to settle a complaint filed by 

an employee with the Department of 

Labor pursuant to section 211 of the 

Energy Reorganization Act of 1974, as 

amended, may contain any provision 

which would prohibit, restrict, or 

otherwise discourage an employee from 

participating in protected activity as 

defined in paragraph (a)(1) of this 

section including, but not limited to, 

providing information to the NRC or to 

his or her employer on potential 

violations or other matters within NRC’s 

regulatory responsibilities. 

(g) Part 19 of 10 CFR sets forth 

requirements and regulatory provisions 

applicable to licensees, holders of a 

standard design approval, applicants for 

a license, standard design certification, 

or standard design approval, and 

contractors or subcontractors of a 

Commission licensee, or holder of a 

standard design approval, and are in 

addition to the requirements in this 

section. 


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§ 53.070 Completeness and accuracy of 

information. 

(a) Information provided to the 

Commission by a holder of a license, 

permit, design certification, or standard 

design approval under this part or an 

applicant for a license, permit, design 

certification, or standard design 

approval under this part, and 

information required by statute or by the 

Commission’s regulations, orders, 

license conditions, or terms and 

conditions of a standard design 

approval to be maintained by the 

applicant or the licensee must be 

complete and accurate in all material 

respects. 

(b) Each applicant or licensee, each 

holder of a standard design approval 

under this part, and each applicant for 

a standard design certification under 

this part following Commission 

adoption of a final design certification 

regulation, must notify the Commission 

of information identified by the 

applicant or licensee as having for the 

regulated activity a significant 

implication for public health and safety 

or common defense and security. An 

applicant, licensee, or holder violates 

this paragraph only if the applicant, 

licensee, or holder fails to notify the 

Commission of information that the 

applicant, licensee, or holder has 

identified as having a significant 

implication for public health and safety 

or common defense and security. 

Notification must be provided to the 

Administrator of the appropriate 

Regional Office within 2 working days 

of identifying the information. This 

requirement is not applicable to 

information which is already required to 

be provided to the Commission by other 

reporting or updating requirements. 

§ 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. 

§ 53.090 Standards for review. 

(a) Common standards. In 

determining that a CP, OL, early site 

permit, COL, or ML under this part will 

be issued to an applicant, the 

Commission will be guided by the 

following considerations: 

(1) Except for an early site permit or 

ML, the processes to be performed, the 

operating procedures, the facility and 

equipment, the use of the facility, and 

other technical specifications, or the 

proposals, in regard to any of the 

foregoing, collectively provide 

reasonable assurance that the applicant 

will comply with the regulations in this 

chapter, including the regulations in 10 

CFR part 20, and that the health and 

safety of the public will not be 

endangered. 

(2) The applicant for a CP, OL, COL, 

or ML is technically and financially 

qualified to engage in the proposed 

activities in accordance with the 

regulations in this chapter. However, no 

consideration of financial qualification 

is necessary for an electric utility 

applicant for an OL for a utilization 

facility of the type described in 

paragraph (d) of this section or for an 

applicant for an ML. 

(3) The issuance of a CP, OL, early site 

permit, COL, or ML to the applicant will 

not, in the opinion of the Commission, 

be inimical to the common defense and 

security or to the health and safety of 

the public. 

(4) Any applicable requirements of 

subpart A of 10 CFR part 51 have been 

satisfied. 

(b) Additional standards for licenses. 

In determining whether a license will be 

issued to an applicant, the Commission 

will, in addition to applying the 

standards set forth in paragraph (a) of 

this section, consider whether the 

proposed activities will serve a useful 

purpose proportionate to the quantities 

of SNM or source material to be utilized. 

(c) Additional standards and 

provisions affecting licenses for 

commercial power. In addition to 

applying the standards set forth in 

paragraphs (a) and (b) of this section, 

paragraphs (c)(1) through (c)(4) of this 

section apply in the case of a license for 

a facility for the generation of 

commercial power. 

(1) The NRC will— 

(i) Give notice in writing of each 

application to the regulatory agency or 

State as may have jurisdiction over the 

rates and services incident to the 

proposed activity; 

(ii) Publish notice of the application 

in trade or news publications as it 


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deems appropriate to give reasonable 

notice to municipalities, private 

utilities, public bodies, and cooperatives 

which might have a potential interest in 

the utilization or production facility; 

and 

(iii) Publish notice of the application 

once each week for four consecutive 

weeks in the Federal Register. No 

license will be issued by the NRC prior 

to the giving of these notices and until 

four weeks after the last notice is 

published in the Federal Register. 

(2) If there are conflicting applications 

for a limited opportunity for such 

license, the Commission will give 

preferred consideration in the following 

order: first, to applications submitted by 

public or cooperative bodies for 

facilities to be located in high cost 

power areas in the United States; 

second, to applications submitted by 

others for facilities to be located in such 

areas; third, to applications submitted 

by public or cooperative bodies for 

facilities to be located in areas other 

than high cost power areas; and, fourth, 

to all other applicants. 

(3) The licensee who transmits 

electric energy in interstate commerce, 

or sells it at wholesale in interstate 

commerce, must be subject to the 

regulatory provisions of the Federal 

Power Act. 

(4) Nothing shall preclude any 

government agency, now or hereafter 

authorized by law to engage in the 

production, marketing, or distribution of 

electric energy, if otherwise qualified, 

from obtaining a CP, OL, or COL under 

this part for a utilization facility for the 

primary purpose of producing electric 

energy for disposition for ultimate 

public consumption. 

(d) Licenses for commercial nuclear 

plants. A license will be issued, to an 

applicant who qualifies, for any one or 

more of the following: to transfer or 

receive in interstate commerce, or 

manufacture, produce, transfer, acquire, 

possess, or use a utilization facility for 

industrial or commercial purposes. 

§ 53.100 Jurisdictional limits. 

No permit, license, standard design 

approval, or standard design 

certification under this part shall be 

deemed to have been issued for 

activities that are not under or within 

the jurisdiction of the United States. 

§ 53.110 Attacks and destructive acts. 

Licensees, applicants for licenses, 

permits, certifications, and design 

approvals, and applicants for an 

amendment to any license, permit, 

certification, or design approval under 

this part are not required to provide for 

design features or other measures for the 

specific purpose of protection against 

the effects of— 

(a) Attacks and destructive acts, 

including sabotage, directed against the 

facility by an enemy of the United 

States, whether a foreign government or 

other person; or 

(b) Use or deployment of weapons 

incident to U.S. defense activities. 

§ 53.115 Rights related to special nuclear 

material. 

(a) No right to the SNM will be 

conferred by a license issued under this 

part except as may be defined by the 

license. 

(b) Neither a license issued under this 

part, nor any right thereunder, nor any 

right to utilize or produce SNM may be 

transferred, assigned, or disposed of in 

any manner, either voluntarily or 

involuntarily, directly or indirectly, 

through transfer of control of the license 

to any person, unless the Commission, 

after securing full information, finds 

that the transfer is in accordance with 

the provisions of the Act and gives its 

consent in writing. 

§ 53.117 License suspension and rights of 

recapture. 

Any license issued under this part 

must be subject to suspension and to the 

rights of recapture of the material or 

control of the facility reserved to the 

Commission under section 108 of the 

Act in a state of war or national 

emergency declared by Congress. 

§ 53.120 Information collection 

requirements: OMB approval. 

(a) The NRC has submitted the 

information collection requirements 

contained in this part to the Office of 

Management and Budget (OMB) for 

approval as required by the Paperwork 

Reduction Act (44 U.S.C. 3501 et seq.). 

The NRC may not conduct or sponsor, 

and a person is not required to respond 

to, a collection of information unless it 

displays a currently valid OMB control 

number. OMB has approved the 

information collection requirements 

contained in this part under control 

number 3150–XXXX. 

(b) The approved information 

collection requirements contained in 

this part appear in §§ 53.070, 53.080, 

53.240, 53.410, 53.420, 53.425, 53.430, 

53.440, 53.450, 53.480, 53.500, 53.540, 

53.605, 53.610, 53.620, 53.700, 53.710, 

53.715, 53.720, 53.730, 53.780, 53.785, 

53.805, 53.810, 53.815, 53.830, 53.850, 

53.855, 53.865, 53.870, 53.875, 53.880, 

53.910, 53.1010, 53.1020, 53.1030, 

53.1045, 53.1060, 53.1070, 53.1075, 

53.1080, 53.1100, 53.1109, 53.1115, 

53.1130, 53.1140, 53.1144, 53.1146, 

53.1173, 53. 1182, 53.1188, 53.1200, 

53.1206, 53.1209, 53.1210, 53.1221, 

53.1230, 53.1236, 53.1239, 53.1241, 

53.1254, 53.1257, 53,1263, 53.1270, 

53.1276, 53.1279, 53.1282, 53.1288, 

53.1295, 53.1300, 53.1306, 53.1309, 

53.1312, 53.1327, 53.1330, 53.1333, 

53.1336, 53.1348, 53.1360, 53.1366, 

53.1369, 53.1372, 53.1384, 53.1410, 

53.1413, 53.1416, 53.1419, 53.1437, 

53.1449, 53.1452, 53.1458, 53.1470, 

53.1505, 53.1510, 53.1515, 53.1525, 

53.1530, 53.1535, 53.1540, 53.1545, 

53.1550, 53.1560, 53.1565, 53.1570, 

53.1575, 53.1580, 53.1620, 53.1630, 

53.1645, 53.1680, 53.1690, 53.1720. 

(c) This part contains information 

collection requirements in addition to 

those approved under the control 

number specified in paragraph (a) of 

this section. The information collection 

requirement and the control numbers 

under which it is approved are as 

follows: 

(1) In §§ 53.765, 53.770, 53.780, and 

53.795, NRC Form 396 is approved 

under control number 3150–0024. 

(2) In §§ 53.775 and 53.795, NRC 

Form 398 is approved under control 

number 3150–0090. 

(3) In § 53.1640, NRC Form 366 is 

approved under control number 3150– 

0104. 

(4) In § 53.1630, NRC Form 361 is 

approved under control number 3150– 

0238. 

(5) In § 53.1650, International Atomic 

Energy Agency Design Information 

Questionnaire forms are approved under 

control number 3150–0056. 

(6) In § 53.1650, DOC/NRC Form AP– 

A and associated forms are approved 

under control numbers 0694–0135. 

Subpart B—Technology-Inclusive 

Safety Requirements 

§ 53.210 Safety criteria for design-basis 

accidents. 

Design features and programmatic 

controls must be provided for each 

commercial nuclear plant such that 

identification and analyses of designbasis accidents (DBAs) in accordance 

with § 53.240 demonstrate the 

following: 

(a) An individual located at any point 

on the boundary of the exclusion area 

for any 2-hour period following the 

onset of the postulated fission product 

release would not receive a radiation 

dose in excess of 25 rem (250 

millisieverts) total effective dose 

equivalent (TEDE); and 

(b) An individual located at any point 

on the outer boundary of the lowpopulation zone who is exposed to the 

radioactive cloud resulting from the 

postulated fission product release 

(during the entire period of its passage) 

would not receive a radiation dose in 


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excess of 25 rem (250 millisieverts) 

TEDE.1

1The use of 25 rem TEDE is not intended 

to imply that this number constitutes an 

acceptable limit for an emergency dose to the 

public under accident conditions. Rather, 

this dose value has been set forth in this 

section as a reference value, which can be 

used in the evaluation of plant design 

features with respect to postulated reactor 

accidents, to assure that these designs 

provide assurance of low risk of public 

exposure to radiation, in the event of an 

accident. 

§ 53.220 Safety criteria for licensing-basis 

events other than design-basis accidents. 

Design features and programmatic 

controls must be provided for each 

commercial nuclear plant such that 

identification and analysis of licensingbasis events (LBEs) other than DBAs in 

accordance with § 53.240 demonstrate 

the following: 

(a) Plant SSCs, personnel, and 

programs provide the necessary 

capabilities and maintain the necessary 

reliability to address LBEs other than 

DBAs in accordance with §§ 53.240 and 

53.450(e), and provide measures for 

defense in depth in accordance with 

§ 53.250; and 

(b) The analysis of risks to public 

health and safety resulting from LBEs 

other than DBAs under § 53.450(e) 

includes comprehensive risk metrics 

that satisfy associated risk performance 

objectives that are acceptable to the NRC 

and provide an appropriate level of 

safety. 

§ 53.230 Safety functions. 

(a) The primary safety function is 

limiting the release of radioactive 

materials from the facility and must be 

maintained during normal operation 

and for LBEs over the life of the plant. 

(b) Additional safety functions needed 

to support the retention of radioactive 

materials during LBEs—such as 

controlling reactivity, heat generation, 

heat removal, and chemical 

interactions—must be identified for 

each commercial nuclear plant. 

(c) The primary and additional safety 

functions are required to satisfy the 

safety criteria defined in §§ 53.210 and 

53.220, or more restrictive alternative 

criteria adopted under § 53.470, and 

must be fulfilled by the design features, 

human actions, and programmatic 

controls specified throughout this part. 

§ 53.240 Licensing-basis events. 

(a) Licensing-basis events must be 

identified for each commercial nuclear 

plant and analyzed under § 53.450 to 

demonstrate that the safety 

requirements in this subpart have been 

satisfied. 

(b) The identified LBEs, ranging from 

anticipated event sequences to very 

unlikely event sequences, must 

collectively address combinations of 

malfunctions of plant SSCs, human 

errors, facility hazards, and the effects of 

external hazards. 

(c) The analysis of LBEs must— 

(1) Include analysis of one or more 

DBAs under § 53.450(f); 

(2) Confirm the adequacy of design 

features and programmatic controls 

needed to satisfy the safety criteria 

defined in §§ 53.210 and 53.220, or 

more restrictive alternative criteria 

adopted under § 53.470, and 

(3) Establish related functional 

requirements for plant SSCs, personnel, 

and programs. 

§ 53.250 Defense in depth. 

(a) Measures must be taken for each 

commercial nuclear plant to ensure 

appropriate defense in depth is 

provided to compensate for 

uncertainties in the analysis of the 

safety criteria such that there is 

reasonable assurance that the safety 

criteria in this subpart are met over the 

life of the plant. 

(b) The uncertainties that must be 

addressed under paragraph (a) of this 

section include those related to the state 

of knowledge and modeling capabilities, 

the ability of barriers to limit the release 

of radioactive materials from the facility 

during LBEs other than DBAs, the 

reliability and performance of plant 

SSCs and personnel, and the 

effectiveness of programmatic controls. 

(c) The safety analysis may not rely 

upon a single engineered design feature, 

human action, or programmatic control, 

no matter how robust, to address the 

range of LBEs other than DBAs. 

§ 53.260 Normal operations. 

Holders of licenses to operate 

commercial nuclear plants under this 

part must control public doses and dose 

rates in unrestricted areas from normal 

plant operations to meet the 

requirements in 10 CFR part 20. 

§ 53.270 Protection of plant workers. 

Holders of licenses to operate 

commercial nuclear plants under this 

part must control occupational doses to 

meet the requirements in 10 CFR part 

20. 

Subpart C—Design and Analysis 

Requirements 

§ 53.400 Design features for licensingbasis events. 

(a) Design features must be provided 

for each commercial nuclear plant such 

that, when combined with 

corresponding human actions and 

programmatic controls, the plant will 

satisfy the safety criteria defined in 

§§ 53.210 and 53.220, or more restrictive 

alternative criteria adopted under 

§ 53.470. 

(b) Design features must ensure that 

the safety functions identified in 

§ 53.230 are fulfilled during licensingbasis events (LBEs). 

§ 53.410 Functional design criteria for 

design-basis accidents. 

(a) Functional design criteria must be 

defined for each design feature required 

by § 53.400 and relied upon to 

demonstrate compliance with the safety 

criteria defined in § 53.210. 

(b) Corresponding human actions and 

programmatic controls must be 

identified and implemented in 

accordance with this and other subparts 

to achieve and maintain the reliability 

and capability of structures, systems, 

and components (SSCs) relied upon to 

satisfy the defined functional design 

criteria and the safety criteria required 

in § 53.210, and to maintain consistency 

with analyses required by § 53.450(f). 

§ 53.415 Protection against external 

hazards. 

Safety-related (SR) SSCs must be 

protected against or must be designed to 

withstand the effects of natural 

phenomena (e.g., earthquakes, 

tornadoes, hurricanes, floods, tsunami, 

and seiches) and constructed hazards 

(e.g., dams, transportation routes, 

military and industrial facilities) 

considering an event severity up to the 

design-basis external hazard levels as 

determined under § 53.510 without 

losing the capability to perform the 

safety functions identified under 

§ 53.230. Specific requirements for 

earthquake engineering are included in 

§ 53.480. 

§ 53.420 Functional design criteria for 

licensing-basis events other than designbasis accidents. 

(a) Functional design criteria must be 

defined for each design feature required 

by § 53.400 and relied upon to— 

(1) Demonstrate compliance with the 

safety criteria in § 53.220 or more 

restrictive alternative criteria adopted 

under § 53.470; and 

(2) Demonstrate compliance with the 

evaluation criteria in § 53.450(e) or more 

restrictive alternative criteria adopted 

under § 53.470. 

(b) Corresponding human actions and 

programmatic controls must be 

identified and implemented in 

accordance with this and other subparts 

to achieve and maintain the reliability 

and capability of SSCs relied upon to— 


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(1) Satisfy the safety criteria in 

§ 53.220 or more restrictive alternative 

criteria adopted under § 53.470; and 

(2) Satisfy the evaluation criteria in 

§ 53.450(e) or more restrictive alternate 

criteria adopted under § 53.470. 

§ 53.425 Design features and functional 

design criteria for normal operations. 

(a) Design features must be provided 

for each commercial nuclear plant to 

support the Radiation Protection 

Program required in § 53.850. 

(b) Functional design criteria must be 

defined for each design feature relied 

upon to demonstrate compliance with 

§ 53.850. 

(c) Functional design criteria, 

including design objectives for dose to 

the maximally exposed member of the 

public, must be defined for design 

features to show that plant design 

features and corresponding 

programmatic controls, including 

monitoring programs, control liquid, 

gaseous, and solid wastes, as required 

under part 20 of this chapter.1

1A guide for keeping doses to the public 

as low as is reasonably achievable is that the 

estimated annual dose to the maximally 

exposed member of the public does not 

exceed 10 mrem total effective dose 

equivalent. A design objective of maintaining 

doses below 10 mrem/year should not be 

construed as a radiation protection standard. 

§ 53.430 Design features and functional 

design criteria for protection of plant 

workers. 

(a) Design features must be provided 

for each commercial nuclear plant such 

that, when combined with 

corresponding programmatic controls, 

the requirements in § 53.270 can be met. 

(b) Functional design criteria must be 

defined for each design feature relied 

upon to demonstrate compliance with 

§ 53.270. 

§ 53.440 Design requirements. 

(a)(1) Analysis, appropriate test 

programs, prototype testing, operating 

experience, or a combination thereof 

must demonstrate that each design 

feature required by § 53.400 meets the 

defined functional design criteria 

required by §§ 53.410 and 53.420. This 

demonstration must consider 

interdependent effects throughout the 

commercial nuclear plant and the range 

of conditions under which the design 

features required by § 53.400 must 

function throughout the plant’s lifetime. 

(2) The design processes for SR and 

non-safety-related but safety-significant 

(NSRSS) SSCs under this part must 

include administrative procedures for 

evaluating operating, design, and 

construction experience and for 

considering applicable important 

industry experiences in the design of 

those SSCs. 

(b) The design features required by 

§ 53.400 must, wherever applicable, be 

designed using generally accepted 

consensus codes and standards that 

have been endorsed or otherwise found 

acceptable by the U.S. Nuclear 

Regulatory Commission (NRC). 

(c) The materials used for each SR and 

NSRSS SSC must be qualified for their 

service conditions over the design life of 

the SSC. 

(d) Possible degradation mechanisms 

related to aging, fatigue, chemical 

interactions, operating temperatures, 

effects of irradiation, and other 

environmental factors that may affect 

the performance of SR and NSRSS SSCs 

must be evaluated and used to inform 

the design and the development of 

integrity assessment programs under 

§ 53.870. 

(e)(1) Safety-related and NSRSS SSCs 

must be designed and located to 

minimize, consistent with other safety 

requirements in this part, the 

probability and effect of fires and 

explosions. 

(2) Noncombustible and fire-resistant 

materials must be used wherever 

practical throughout the facility, 

particularly in locations with SR and 

NSRSS SSCs. 

(3) Fire detection and fire suppression 

systems of appropriate capacity and 

capability must be provided and 

designed to minimize the adverse effects 

of fires on SR and NSRSS SSCs. 

(4) Fire suppression systems must be 

designed to ensure that their rupture or 

inadvertent operation does not 

significantly impair the ability of SR 

and NSRSS SSCs to perform their safety 

functions to satisfy § 53.230. 

(f) Safety and security must be 

considered together in the design 

process such that, where possible, 

security issues are effectively resolved 

through design and engineered security 

features. 

(g) The reactor system and waste 

stores for each commercial nuclear plant 

must be capable of achieving and 

maintaining a subcritical condition 

during normal operations and following 

any LBE identified in accordance with 

§ 53.240. 

(h) Each commercial nuclear plant 

must have a capability to provide longterm cooling of the reactor fuel and 

waste stores during normal operations 

and following any LBE identified in 

accordance with § 53.240. 

(i) The design, analysis, staffing, and 

programmatic controls for each 

commercial nuclear plant must consider 

the number of reactors, waste stores, 

and other significant inventories of 

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

1Changes to the detailed parameters on 

aircraft impact characteristics set forth in 

guidance must be approved by the 

Commission. 

(k) Design features and related 

functional design criteria must be 

defined such that analyses demonstrate 

a low risk of permanent injury to the 

public due to the health effects of the 

chemical hazards of licensed material. 

(l) Measures must be taken during the 

design of commercial nuclear plants to 

minimize, to the extent practicable, 

contamination of the facility and the 

environment, facilitate eventual 

decommissioning, and minimize, to the 

extent practicable, the generation of 

radioactive waste in accordance with 

§ 20.1406 of this chapter. 

(m)(1) Each commercial nuclear plant 

must include criticality monitoring 

capabilities meeting the requirements of 

either § 70.24 of this chapter or 

paragraph (m)(2) of this section. 

(2) In lieu of maintaining a monitoring 

system capable of detecting criticality as 

described in § 70.24 of this chapter, 

criticality accident requirements may be 

satisfied by— 

(i) Demonstrating the sub-criticality of 

special nuclear material, except when it 

is inside the reactor and the reactor is 

being operated, by maintaining keffective below 0.95 at a 95 percent 

probability, 95 percent confidence level, 

under conditions that maximize 

reactivity for the applicable storage and 

handling configurations, and 

(ii) Providing radiation monitors for 

fuel storage and associated handling 


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areas when fuel is present to detect 

excessive radiation levels and to 

support initiating appropriate safety 

actions. 

(3) While a spent fuel transportation 

package approved under 10 CFR part 71 

of this chapter or spent fuel storage cask 

approved under 10 CFR part 72 is in the 

special nuclear material handing or 

storage area, the requirements in 10 CFR 

parts 71 or 72, as applicable, and the 

requirements of the certificate of 

compliance for that package or cask, are 

the applicable requirements for the fuel 

within that package or cask. 

(n)(1) The design of each commercial 

nuclear plant must reflect state-of-theart human factors principles for safe and 

reliable performance in all locations that 

human activities are expected for 

performing or supporting the continued 

availability of plant safety or emergency 

response functions. 

(2) The design must provide for the 

capabilities described in § 53.730(b) to 

ensure the plant staff are able to monitor 

plant conditions and respond to events. 

(3) The means by which the design 

and human actions together will achieve 

the safety requirements of subpart B of 

this part must be evaluated and used to 

inform the design and the development 

of the concept of operations required by 

§ 53.730(c). 

(4) A functional requirements analysis 

and function allocation must be used to 

ensure that plant design features 

address how safety functions and 

functional safety criteria are satisfied, 

and how the safety functions will be 

assigned to appropriate combinations of 

human action, automation, active safety 

features, passive safety features, or 

inherent safety characteristics. 

§ 53.450 Analysis requirements. 

(a) Requirement to have a 

probabilistic risk assessment (PRA). A 

PRA of each commercial nuclear plant 

must be performed to identify potential 

failures, susceptibility to internal and 

external hazards, and other contributing 

factors to event sequences that might 

challenge the safety functions identified 

in § 53.230 and to support 

demonstrating that each commercial 

nuclear plant meets the safety criteria of 

§ 53.220, or more restrictive alternative 

criteria adopted under § 53.470. 

(b) Specific uses of analyses. The PRA 

in combination with other generally 

accepted approaches for systematically 

evaluating engineered systems must be 

used— 

(1) In informing the selection of the 

LBEs, as described in § 53.240, which 

must be considered in the design to 

determine compliance with the safety 

criteria in subpart B of this part. 

(2) For informing the classification of 

SSCs according to their safety 

significance in accordance with § 53.460 

and for identifying the environmental 

conditions under which the SSCs and 

operating staff must perform their safety 

functions. 

(3) In evaluating the adequacy of 

defense-in-depth measures required in 

accordance with § 53.250. 

(4) To identify and assess all plant 

operating states where there is the 

potential for the uncontrolled release of 

radioactive material to the environment. 

(5) To identify and assess events that 

challenge plant control and safety 

systems whose failure could lead to the 

uncontrolled release of radioactive 

material to the environment. These 

include internal events, such as human 

errors and equipment failures, and 

external events identified in accordance 

with subpart D of this part. 

(c) Maintenance and upgrade of 

analyses. The PRA must be maintained 

at least every 5 years until the 

permanent cessation of operations 

under § 53.1070 and upgraded in 

conformance with generally accepted 

methods, standards, and practices that 

have been endorsed or otherwise found 

acceptable by the NRC. 

(d) Qualification of analytical codes. 

The analytical codes used in modeling 

plant behavior in analyses of licensingbasis events (including but not limited 

to thermodynamics, reactor physics, 

fuel performance, and mechanistic 

source term codes) must be qualified for 

the range of conditions for which they 

are to be used. 

(e) Analyses of licensing-basis events 

other than design-basis accidents. 

(1) Analyses must be performed for 

LBEs other than design-basis accidents 

(DBAs). These LBEs must be identified 

using insights from a PRA in 

combination with other generally 

accepted approaches for systematically 

evaluating engineered systems to 

identify and analyze equipment failures 

and human errors. 

(2) The analysis of LBEs other than 

DBAs must include definition of 

evaluation criteria for each event or 

specific categories of LBEs to determine 

the acceptability of the plant response to 

the challenges posed by internal and 

external hazards to provide an 

appropriate level of safety. 

(3) The analyses of LBEs other than 

DBAs must address event sequences 

from initiation to a defined end state 

and be used in combination with other 

engineering analyses to demonstrate 

that the functional design criteria 

required by § 53.420 provide sufficient 

barriers to the unplanned release of 

radionuclides to satisfy the evaluation 

criteria defined for each LBE other than 

DBAs, to satisfy the safety criteria 

specified in accordance with § 53.220 

and provide defense in depth as 

required by § 53.250. 

(4) The methodology used to identify, 

categorize, and analyze LBEs must 

include a means to identify event 

sequences deemed significant for 

controlling the risks posed to public 

health and safety. 

(f) Analysis of design-basis accidents. 

(1) The analysis of LBEs required by 

§ 53.240 must include analysis of DBAs 

that address possible challenges to the 

safety functions identified under 

§ 53.230. The events selected as DBAs 

must be those that, if not terminated, 

have the potential for exceeding the 

safety criteria in § 53.210. 

(2) The DBAs selected must be 

analyzed using deterministic methods 

that address event sequences from 

initiation to a safe stable end state and 

assume only the SR SSCs identified 

under § 53.460 and human actions 

addressed by the requirements of 

subpart F of this part are available to 

perform the safety functions identified 

in accordance with § 53.230. 

(3) The analysis must conservatively 

demonstrate compliance with the safety 

criteria in § 53.210. 

(g) Other required analyses. Analyses 

must be performed to assess— 

(1) Fire protection. Fire protection 

measures to demonstrate, through 

inclusion of fires in the analysis of LBEs 

or by separate analyses, that a fire or 

explosion in any plant area would not— 

(i) Prevent equipment from fulfilling 

the safety functions identified in 

accordance with § 53.230, or 

(ii) Challenge the safety criteria in 

§§ 53.210 and 53.220. 

(2) Aircraft impact. Measures 

provided to protect against aircraft 

impacts under § 53.440(j). 

(3) Dose to members of the public. 

Measures taken under § 53.425, 

including estimating— 

(i) The quantity of each of the 

principal radionuclides expected to be 

released annually to unrestricted areas 

in liquid effluents produced during 

normal reactor operations and the dose 

to the maximally exposed member of 

the public in unrestricted areas. 

(ii) The quantities of each of the 

principal radionuclides of the gases, 

halides, and particulates expected to be 

released annually to unrestricted areas 

in gaseous effluents produced during 

normal reactor operations and the dose 

to the maximally exposed member of 

the public in unrestricted areas. 

(iii) The annual external radiation 

dose in unrestricted areas and the 

maximally exposed member of the 


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public in unrestricted areas due to 

direct radiation from contained 

radiation sources from the commercial 

nuclear plant during normal reactor 

operations. 

§ 53.460 Safety categorization and special 

treatments. 

(a) Structures, systems, and 

components must be classified 

according to their safety significance. 

The SSC categories must include 

‘‘Safety-Related,’’ ‘‘Non-Safety-Related 

but Safety-Significant,’’ and ‘‘NonSafety-Significant,’’ as defined in 

subpart A of this part. 

(b) For SR and NSRSS SSCs, the 

conditions under which they must 

perform their safety function in § 53.230 

must be identified. Special treatments 

must be established in accordance with 

this and other subparts to provide 

confidence that the SSCs will perform 

under the service conditions and with 

reliability consistent with the analysis 

performed under § 53.450 to 

demonstrate meeting the safety criteria 

in §§ 53.210 and 53.220, or more 

restrictive alternative criteria adopted 

under § 53.470. 

(1) The special treatments for SR SSCs 

must include meeting the applicable 

quality assurance requirements from 

appendix B of part 50 of this chapter. 

(2) The special treatments for NSRSS 

SSCs and special treatments for SR SSCs 

beyond those required under (b)(1) of 

this section may include meeting 

selected quality assurance requirements 

from appendix B of part 50 of this 

chapter when such treatment is needed 

to address performance requirements, 

equipment reliability, or uncertainties. 

(c) Human actions needed to prevent 

or mitigate LBEs must be identified, be 

able to be performed reliably under the 

postulated environmental conditions, 

and be addressed by programs 

established in accordance with subpart 

F of this part to provide confidence that 

those actions will be performed as 

assumed in the analysis performed in 

accordance with § 53.450 to 

demonstrate meeting the criteria in 

§§ 53.210, 53.220, and 53.450(e), or 

more restrictive alternative criteria 

adopted under § 53.470. 

§ 53.470 Maintaining analytical safety 

margins used to justify operational 

flexibilities. 

Where an applicant or licensee so 

chooses, alternative criteria more 

restrictive than those defined in 

§§ 53.220 and 53.450(e) may be adopted 

to support operational flexibilities. In 

such cases, applicants and licensees 

must ensure that the functional design 

criteria of § 53.420, the analysis 

requirements of § 53.450(e), and 

identification of special treatment of 

SSCs and human actions under § 53.460 

reflect and support the use of alternative 

criteria to justify operational 

flexibilities. Licensees must ensure that 

measures taken to provide the analytical 

margins supporting operational 

flexibilities are incorporated into design 

features and programmatic controls and 

are maintained within programs 

required in other subparts. 

§ 53.480 Earthquake engineering. 

(a) Effects of earthquakes. Structures, 

systems, and components classified as 

SR or NSRSS must be able to withstand 

the effects of earthquakes, 

commensurate with the safety 

significance of the SSC, without loss of 

capability to perform their role in 

fulfilling the safety functions required 

by § 53.230. 

(b) Definitions. For the purpose of this 

section— 

Design-Basis Ground Motions 

(DBGMs) are the vibratory ground 

motions for which certain SSCs must be 

designed to remain functional. 

Operating basis earthquake (OBE) 

ground motion is the vibratory ground 

motion for which those features of the 

commercial nuclear plant necessary for 

continued operation without undue risk 

to the health and safety of the public are 

designed to remain functional. The OBE 

ground motion is used in § 53.720. 

Response spectrum is a plot of the 

maximum responses (acceleration, 

velocity, or displacement) of idealized 

single-degree-of-freedom oscillators as a 

function of the natural frequencies of 

the oscillators for a given damping 

value. The response spectrum is 

calculated for a specified vibratory 

motion input at the oscillators’ 

supports. 

Surface deformation is the distortion 

of geologic strata on or near the ground 

surface that occurs because of tectonic 

forces that result from earthquakes. 

(c) Design considerations—(1) DesignBasis Ground Motions. (i) The DBGMs 

must be derived from the Site Ground 

Motion Response Spectra developed in 

accordance with § 53.510(c), by taking 

into consideration the functional design 

criteria of SSCs in accordance with 

§§ 53.410 and 53.420. The horizontal 

component of the DBGM(s) in the freefield at the foundation level of the 

structures must be an appropriate 

response spectrum that is determined 

based on the risk-significance of SSCs 

and their safety functions. In view of the 

limited data available on vibratory 

ground motion of strong earthquakes, it 

is acceptable that the design response 

spectra be smoothed spectra. 

(ii) The commercial nuclear plant 

must be designed so that, if the DBGMs 

occur, the following SSCs remain 

functional and within applicable stress, 

strain, and deformation limits: 

(A) Structures, systems, and 

components for which functional design 

criteria are established in accordance 

with § 53.410 or § 53.420; and 

(B) Structures, systems, and 

components classified as SR or NSRSS 

commensurate with safety significance 

in accordance with § 53.460. 

(iii) In addition to seismic loads, 

applicable concurrent normal operating, 

functional, and accident-induced loads 

must be taken into account in the design 

of the SR SSCs and, commensurate with 

safety significance, NSRSS SSCs. 

(iv) The design of the commercial 

nuclear plant must take into account the 

possible effects of seismic-induced 

ground disruption, such as fissuring, 

lateral spreads, differential settlement, 

liquefaction, and landsliding, on the 

facility foundations. 

(v) The SSCs fulfilling the safety 

functions required by § 53.230 must be 

demonstrated through design, testing, or 

qualification methods to be able to 

fulfill those safety functions during and 

after the vibratory ground motion 

associated with the DBGMs. 

(vi) The evaluation of SSCs required 

by this section to show they are able to 

function during and after earthquake 

ground motion must take into account 

soil-structure interaction effects and the 

expected duration of vibratory motion. 

It is permissible to design for strain 

limits in excess of yield strain in some 

of these SSCs during the DBGMs and 

under the postulated concurrent loads, 

provided the necessary safety functions 

are maintained. 

(2) OBE Ground Motion. The OBE 

Ground Motion must be characterized 

by response spectra. The value of the 

OBE Ground Motion must be set to onethird or less of the DBGMs response 

spectra. 

(3) [Reserved] 

(4) Required seismic instrumentation. 

Suitable instrumentation must be 

provided so that the seismic response of 

commercial nuclear plant SR SSCs or 

NSRSS SSCs can be evaluated promptly 

after an earthquake. 

(d) Surface deformation. (1) The 

potential for surface deformation must 

be taken into account in the design of 

the commercial nuclear plant by 

providing reasonable assurance that in 

the event of deformation, SSCs 

classified as SR or NSRSS in accordance 

with § 53.460 will remain functional. 

(2) In addition to surface deformation 

induced loads, the design of SSCs must 

take into account, commensurate with 


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safety significance, seismic loads and 

applicable concurrent functional and 

accident-induced loads. 

(3) The design provisions for surface 

deformation must be based on its 

postulated occurrence in any direction 

and azimuth and under any part of the 

commercial nuclear plant, unless 

evidence indicates this assumption is 

not appropriate, and must take into 

account the estimated rate at which the 

surface deformation may occur. 

(e) Seismically induced floods and 

water waves and other design 

conditions. Seismically induced floods 

and water waves from either locally or 

distantly generated seismic activity and 

other design conditions determined 

pursuant to subpart D of this part must 

be taken into account in the design of 

the commercial nuclear plant so as to 

prevent undue risk to the health and 

safety of the public. 

(f) Analysis. The analyses required by 

§ 53.450 must address seismic hazards 

and related SSC responses in 

determining that the safety criteria 

defined in § 53.220 will be met. 

(g) Design criteria, human actions, 

and programmatic controls. Functional 

design criteria, human actions, and 

programmatic controls needed to 

address seismic events must be 

identified and implemented in 

accordance with this and other subparts 

to achieve and maintain the 

performance of SSCs relied upon to 

satisfy the safety criteria in § 53.220 and 

to maintain consistency with analyses 

required by § 53.450 when accounting 

for the site-specific frequencies and 

magnitudes of earthquakes for a 

commercial nuclear plant. 

Subpart D—Siting Requirements 

§ 53.500 General siting and siting 

assessment. 

(a) The siting of each commercial 

nuclear plant must be supported by 

assessments of proposed sites such that 

the design, including design features 

and programmatic controls 

corresponding to the site characteristics, 

satisfies the safety criteria defined in 

§§ 53.210 and 53.220 or more restrictive 

alternative criteria adopted under 

§ 53.470. The siting assessment must 

ensure that site characteristics that 

might contribute to the initiation, 

progression, or consequences of 

licensing-basis events (LBEs) analyzed 

under §§ 53.450 and 53.480 are 

identified and mitigated by design 

features or programmatic controls. The 

siting assessment must take into 

consideration the potential adverse 

impacts that a commercial nuclear plant 

may have on nearby populations as a 

result of normal operations or LBEs. 

(b) Activities performed to identify 

site characteristics or otherwise needed 

to determine site-specific contributors to 

functional design criteria or analysis 

assumptions under subpart C of this 

part must satisfy the applicable special 

treatment requirements of § 53.460, 

including, where applicable, the quality 

assurance requirements from appendix 

B of part 50 of this chapter. 

§ 53.510 External hazards. 

(a) General external hazard 

requirements. The design-basis external 

hazard level for the relevant external 

hazards for a site must be identified and 

characterized based on site-specific 

assessments of natural and constructed 

hazards with the potential to adversely 

affect plant functions. The external 

hazard frequencies and magnitudes 

determined from the site-specific 

assessments must take into account 

uncertainties and variabilities in data, 

models, and methods relied on to 

characterize the external hazards. 

(b) Definitions. For the purpose of this 

section, the following terms mean: 

Geological siting factors are geological 

and seismic factors that may affect the 

design and operation of the proposed 

commercial nuclear plant. 

Ground Motion Response Spectra 

(GMRS) are the site-specific GMRS 

resulting from the geologic 

investigations and evaluations of the 

site vicinity and region and used to 

determine design-basis ground motions 

for structures, systems, and components 

under § 53.480. 

Probabilistic seismic hazard analysis 

is an analytical methodology that 

incorporates uncertainty into estimates 

of an annual frequency of exceedance 

for a certain ground motion parameter 

(e.g., peak ground acceleration, peak 

ground velocity, response spectral 

values) at a site. 

(c) Geological investigations. The 

GMRS for the site must be determined 

based on the results of investigations of 

the geological, seismological, and 

engineering characteristics of the site 

and its environs and must be 

characterized by both horizontal and 

vertical free-field GMRS at the free 

ground surface. The size of the region to 

be investigated and the type of data 

pertinent to the investigations must be 

determined based on the nature of the 

region surrounding the site. Data on 

vibratory ground motion, earthquake 

recurrence rates, fault geometry and slip 

rates, and site subsurface material 

properties must be obtained by 

reviewing pertinent literature and 

carrying out field investigations. 

Uncertainties are inherent in the 

parameters and models used to estimate 

the GMRS for the site. The site 

assessment must reflect these 

uncertainties through an appropriate 

analysis, such as a probabilistic seismic 

hazard analysis. 

(d) Geologic and seismic siting 

factors. The geologic and seismic siting 

factors considered for design under 

§§ 53.415 and 53.480 must include, but 

are not limited to, determination of the 

potential for surface tectonic and 

nontectonic deformations, the size and 

character of seismically induced floods 

and water waves that could affect a site 

from either locally or distantly 

generated seismic activity, soil and rock 

stability, liquefaction potential, and 

natural and artificial slope stability. 

§ 53.520 Site characteristics. 

Site characteristics that might 

contribute to the initiation, progression, 

or consequences of LBEs analyzed 

under § 53.450 must be identified, 

assessed, and considered in the design 

and analyses required by subpart C of 

this part. 

§ 53.530 Population-related 

considerations. 

Every site must have an exclusion 

area, a low-population zone, and a 

population center distance as defined in 

§ 53.020. 

(a) The offsite radiological 

consequences estimated by the analyses 

required by § 53.450(f) must be used to 

confirm that— 

(1) An individual located at any point 

on the boundary of the exclusion area 

for any 2-hour period following onset of 

the postulated fission product release 

would not receive a radiation dose in 

excess of 25 rem (250 millisieverts) total 

effective dose equivalent. 

(2) An individual located at any point 

on the outer boundary of the lowpopulation zone who is exposed to the 

radioactive cloud resulting from the 

postulated fission product release 

(during the entire period of its passage) 

would not receive a radiation dose in 

excess of 25 rem (250 millisieverts) total 

effective dose equivalent. 

(b) The population center distance 

must be at least one and one-third times 

the distance from the reactor to the 

outer boundary of the low-population 

zone. The boundary of the population 

center must be determined upon 

consideration of population 

distribution. Political boundaries are not 

controlling in the calculation of 

population center distance. 

(c) Reactor sites should be located 

away from very densely populated 

centers. Areas of low-population density 


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are, generally, preferred. However, in 

determining the acceptability of a 

particular site located away from a very 

densely populated center but not in an 

area of low-population density, 

consideration will be given to safety, 

environmental, economic, or other 

factors, which may result in the site 

being found acceptable. 

§ 53.540 Siting interfaces. 

Site characteristics must be addressed 

by the design features, programmatic 

controls, and supporting analyses used 

to demonstrate that the safety criteria in 

§§ 53.210 and 53.220 are met for each 

commercial nuclear plant. Site 

characteristics must be such that 

adequate emergency plans and security 

plans can be developed and maintained. 

Subpart E—Construction and 

Manufacturing Requirements 

§ 53.600 Construction and 

manufacturing—scope and purpose. 

This subpart applies to those 

construction and manufacturing 

activities authorized by a construction 

permit (CP), combined license (COL), 

manufacturing license (ML), or limited 

work authorization (LWA) issued under 

this part. 

§ 53.605 Reporting of defects and 

noncompliance. 

Each CP and ML issued under this 

part is subject to the terms and 

conditions in this section, and each COL 

issued under this part is subject to the 

terms and conditions in this section 

until the date that the Commission 

makes the finding under § 53.1452(g). 

(a) Definitions. The definitions in 

§ 21.3 of this chapter apply to this 

section. 

(b) Posting requirements. (1) Each 

individual, partnership, corporation, 

dedicating entity, or other entity subject 

to the regulations in this section must 

post current copies of this section and 

the regulations in 10 CFR part 21; 

section 206 of the Energy 

Reorganization Act of 1974, as 

amended; and procedures adopted 

under these regulations. These 

documents must be posted in a 

conspicuous position on any premises 

within the United States where the 

activities subject to the license are 

conducted. 

(2) If posting of these regulations or 

the procedures adopted under them is 

not practical, the licensee may, in 

addition to posting section 206 of the 

Energy Reorganization Act of 1974, as 

amended, post a notice that describes 

the regulations/procedures, including 

the name of the individual to whom 

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. 


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(e) Notification—timing and where 

sent. The notification required by 

paragraph (d) of this section must 

consist of— 

(1) Initial notification by telephone, 

facsimile, or email identified in 

appendix A to 10 CFR part 73 to the 

U.S. Nuclear Regulatory Commission 

(NRC) Operations Center within 2 days 

following receipt of information by the 

director or responsible corporate officer 

under paragraph (c)(3) of this section, 

on the identification of a defect or a 

failure to comply. If the CP, COL, or ML 

holder elects to use facsimile, 

verification that the facsimile has been 

received should be made by calling the 

NRC Operations Center. This paragraph 

does not apply to interim reports 

described in paragraph (c)(2) of this 

section. 

(2) Written notification submitted to 

the NRC Document Control Desk by an 

appropriate method listed in § 53.040, 

with a copy to the appropriate NRC 

Regional Administrator at the address 

specified in appendix D to 10 CFR part 

20 and a copy to the appropriate NRC 

resident inspector, if applicable, within 

30 days following receipt of information 

by the director or responsible corporate 

officer under paragraph (c)(3) of this 

section, on the identification of a defect 

or failure to comply. 

(f) Content of notification. The written 

notification required by paragraph (e)(2) 

of this section must clearly indicate that 

the written notification is being 

submitted under this section and 

include the following information, to 

the extent known. 

(1) Name and address of the 

individual or individuals informing the 

Commission. 

(2) Identification of the facility, the 

activity, or the basic component 

supplied for the facility or the activity 

within the United States which contains 

a defect or fails to comply. 

(3) Identification of the firm 

constructing or manufacturing the 

facility or supplying the basic 

component which fails to comply or 

contains a defect. 

(4) Nature of the defect or failure to 

comply and the safety hazard which is 

created or could be created by the defect 

or failure to comply. 

(5) The date on which the information 

of a defect or failure to comply was 

obtained. 

(6) In the case of a basic component 

that contains a defect or failure to 

comply, the number and location of 

these components in use at the facility 

subject to the regulations in this part. 

(7) In the case of a completed reactor 

manufactured under this part, the 

entities to which the reactor was 

supplied. 

(8) The corrective action which has 

been, is being, or will be taken; the 

name of the individual or organization 

responsible for the action; and the 

length of time that has been or will be 

taken to complete the action. 

(9) Any advice related to the defect or 

failure to comply about the facility, 

activity, or basic component that has 

been, is being, or will be given to other 

entities. 

(g) Procurement documents. Each 

holder of a CP, COL, or ML subject to 

this section must ensure that each 

procurement document for a facility or 

a basic component specifies the 

provisions of 10 CFR part 21 or this 

section that apply, as applicable. 

(h) Coordination with 10 CFR part 21. 

The requirements of this section are 

satisfied when the defect or failure to 

comply associated with a substantial 

safety hazard has been previously 

reported under 10 CFR part 21, under 

§ 73.1205 of this chapter, under this 

section, or under § 53.1640. 

(i) Records retention. The holder of a 

CP, COL, or ML subject to this section 

must prepare and maintain records 

necessary to accomplish the purposes of 

this section, specifically— 

(1) Retain procurement documents, 

which define the requirements that 

facilities or basic components must 

satisfy in order to be considered 

acceptable, for the lifetime of the facility 

or basic component. 

(2) Retain records of evaluations of all 

deviations and failures to comply under 

paragraph (c)(1) of this section for the 

longest of— 

(i) Ten years from the date of the 

evaluation; 

(ii) Five years from the date that an 

early site permit is referenced in an 

application for a COL; or 

(iii) Five years from the date of 

delivery of a manufactured reactor. 

(3) Retain records of all interim 

reports to the Commission made under 

paragraph (c)(2) of this section, or 

notifications to the Commission made 

under paragraph (d) of this section for 

the minimum time periods stated in 

paragraph (i)(2) of this section; 

(4) Suppliers of basic components 

must retain records of— 

(i) All notifications sent to affected 

licensees or purchasers under paragraph 

(d)(4) of this section for a minimum of 

10 years following the date of the 

notification; 

(ii) The facilities or other purchasers 

to whom the basic components or 

associated services were supplied for a 

minimum of 15 years from the delivery 

of the basic component or associated 

services. 

(5) Maintaining reports in accordance 

with this section satisfies the 

recordkeeping obligations under 10 CFR 

part 21 of the entities, including 

directors or responsible officers thereof, 

subject to this section. 

§ 53.610 Construction. 

(a) Management and control. 

Licensees must ensure that the 

following plans, programs, and 

organizational units are developed and 

implemented to manage and control the 

construction activities: 

(1) Programs to ensure that the 

construction of a commercial nuclear 

plant supports the eventual compliance 

with the design and analysis 

requirements in subpart C of this part. 

(2) An organization, headed by 

qualified personnel, responsible for 

managing, controlling, and evaluating 

the adequacy of the construction 

activities. 

(3) Procedures describing the 

qualifications for personnel in key 

positions in the licensee’s management 

and control organization and the 

organizational responsibilities, 

authority, and interfaces with other 

parts of the licensee’s organization. 

(4) Procedures to evaluate the 

applicability of other national and 

international construction experience to 

the planned and ongoing construction 

activities and to ensure the applicable 

experience will be provided to those 

constructing the plant. 

(5) A fitness-for-duty program, under 

10 CFR part 26. 

(6)(i) A QAP meeting the 

requirements of appendix B of part 50 

of this chapter as required by 

§ 53.460(b). 

(ii) Appropriate programmatic 

controls to provide special treatment for 

non-safety-related but safety-significant 

structures, systems, and components 

(SSCs). 

(7) A radiation protection program, in 

accordance with 10 CFR part 20, that 

includes measures for monitoring the 

dose to individuals working with 

radioactive materials brought onto the 

site, as applicable. 

(8) An information security program 

in accordance with §§ 73.21, 73.22, and 

73.23 of this chapter, as applicable. 

(b) Construction activities. No person 

may begin the construction of a 

commercial nuclear plant on a site on 

which the facility is to be operated 

under this part until that person has 

been issued either a CP or COL, an early 

site permit authorizing activities under 

§ 53.1130, or an LWA under this part. 

(1) Licensees must satisfy the 

following requirements: 


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(i) As appropriate, considering the 

types and quantities of radioactive 

materials being brought onto the site— 

(A) The licensee must maintain and 

follow a special nuclear material (SNM) 

material control and accounting 

program, a measurement control 

program, and other material control 

procedures that include corresponding 

record management requirements as 

required by the provisions of § 70.32 of 

this chapter. Prior to initial receipt of 

SNM onsite, the licensee must 

implement an SNM material control and 

accounting program in accordance with 

10 CFR part 74. 

(B) Procedures must be in place to 

receive, possess, use, and store source, 

byproduct, and SNM in accordance with 

applicable portions of 10 CFR parts 30, 

40, and 70. 

(C) A plant staff training program 

associated with the receipt of 

radioactive material must be approved 

and implemented prior to initial receipt 

of byproduct, source or SNM (excluding 

exempt quantities as described in 

§ 30.18 of this chapter). 

(ii) For construction of a commercial 

nuclear plant involving multiple reactor 

units, plans and procedures must be in 

place to prevent or mitigate potential 

hazards to the SSCs of operating units 

resulting from construction activities, 

including the managerial and 

administrative controls to be used to 

provide assurance that the limiting 

conditions for operation of the operating 

units are not exceeded as a result of 

construction activities. 

(iii) Procedures must be in place prior 

to the start of construction activities that 

describe how construction will be 

controlled so as not to impact other 

features important to the design, such as 

dewatering, slope stability, backfill, 

compaction, and seepage. 

(iv) For LWA holders, a plan must be 

developed for redress of activities 

performed under the LWA should one 

of the following situations arise: 

(A) LWA work activities are 

terminated by the holder of the LWA; 

(B) The LWA is revoked by the NRC; 

or 

(C) The Commission denies the 

associated CP or COL application. 

(2)(i) Onsite fresh fuel must be 

protected and stored in compliance with 

§ 73.67 of this chapter. 

(ii) Before initial fuel load into the 

reactor (or, for a fueled manufactured 

reactor, before initiating the physical 

removal of any one of the independent 

physical mechanisms to prevent 

criticality required under 

§ 53.620(d)(1)), a cybersecurity program 

that meets the requirements of §§ 73.54 

or 73.110 of this chapter, a physical 

security program that meets the 

requirements of §§ 73.55 or 73.100 of 

this chapter, and an access 

authorization program that meets the 

requirements of §§ 73.56 or 73.120 of 

this chapter must be established, as 

applicable. 

(iii) Fire protection measures must be 

implemented for work and storage areas 

(including adjacent fire areas that could 

affect the work or storage area) before 

initial receipt of byproduct, source, or 

non-fuel SNM (excluding exempt 

quantities as described in § 30.18 of this 

chapter). The fire protection measures 

for areas associated with new fuel 

(including all fuel handling, fuel 

storage, and adjacent fire areas that 

could affect the new fuel) must be 

implemented before receipt of fuel. 

Prior to the receipt of fuel, a formal 

letter of agreement must be in place 

with the local fire department 

specifying the nature of arrangements in 

support of the fire protection program. 

(c) Inspection and acceptance. (1) The 

licensee must have a process for 

accepting individual or groups of SSCs 

upon completion of construction and 

protecting them from damage or 

tampering as other construction 

activities continue. 

(2) The post construction acceptance 

process must address the inspections, 

tests, analyses, and acceptance criteria 

specified in the COL under § 53.1440 or 

the equivalent verifications needed to 

support the issuance of an operating 

license under § 53.1387. 

§ 53.620 Manufacturing. 

(a) Management and control. Holders 

of MLs must ensure that the following 

plans, programs, and organizational 

units are developed and implemented to 

manage and control the manufacturing 

activities within the scope of the ML: 

(1) Programs to ensure that the 

manufacturing of a manufactured 

reactor or portions of a manufactured 

reactor complies with the design and 

analysis requirements in subpart C of 

this part. The entity with design 

authority for the manufactured reactor 

covered by the ML must be identified in 

the license. 

(2) An organizational and 

management structure responsible for 

managing, controlling, and evaluating 

the adequacy of the reactor design and 

manufacturing activities. 

(3) Procedures describing the 

qualifications for personnel in key 

positions in the licensee’s management 

and control organization and the 

organizational responsibilities, 

authority, and interfaces with other 

parts of the licensee’s organization. 

(4) A program to evaluate the 

applicability of other national and 

international design and manufacturing 

experience to the planned and ongoing 

manufacturing activities. 

(5) A fitness-for-duty program, in 

accordance with 10 CFR part 26. 

(6)(i) A QAP meeting the 

requirements of appendix B to part 50 

of this chapter, to be applied to the 

design, fabrication, construction, and 

testing of the SSCs of the manufactured 

reactor. 

(ii) Appropriate programmatic 

controls to provide special treatment 

measures for non-safety-related but 

safety-significant SSCs. 

(7) A radiation protection program, in 

accordance with 10 CFR part 20, that 

includes measures for monitoring the 

dose to individuals if the manufacturing 

activities include working with 

radioactive materials. 

(8) An information security program 

in accordance with §§ 73.21, 73.22 and 

73.23 of this chapter, as applicable. 

(b) Manufacturing activities. Holders 

of MLs must satisfy the following 

requirements: 

(1) The manufacturing process must 

be conducted within facilities for which 

the ML holder has the authority to 

establish controls on any activity that 

might affect manufacturing. The 

licensee must establish access controls 

to the portions of each facility involved 

in the manufacturing processes 

governed by the ML. 

(2) Manufacturing processes must be 

performed in accordance with the ML 

and the referenced codes and standards 

that have been endorsed or otherwise 

found acceptable by the NRC. 

(3) A post-manufacturing inspection 

and acceptance process must be 

established and implemented before 

transporting a manufactured reactor or 

portions of a manufactured reactor for 

installation at a commercial nuclear 

plant. The process must consider the 

results of inspections, tests, and 

analyses that have been performed and 

the acceptance criteria that are 

necessary and sufficient to conclude 

that manufacturing activities have been 

completed in accordance with the ML. 

(c) Control of radioactive materials. 

As appropriate considering the types 

and quantities of radioactive materials 

being brought into the manufacturing 

facility— 

(1) Procedures must be in place to 

receive, transfer, possess, and use 

source, byproduct, and SNM in 

accordance with the applicable portions 

of 10 CFR parts 30, 40 and 70. 

(2) A fire protection program must be 

established and implemented before the 

initial receipt of byproduct, source, or 


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non-fuel SNM (excluding exempt 

quantities as described in § 30.18 of this 

chapter). 

(3) An emergency plan appropriate for 

responding to the facility-specific 

hazards of an accidental release of 

radioactive material and to limit the 

health effects of the associated chemical 

hazards of licensed material must be 

approved and implemented prior to the 

receipt of byproduct, source, or SNM 

(excluding exempt quantities as 

described in § 30.18 of this chapter). 

(4) A plant staff training program 

associated with the receipt of 

radioactive material must be approved 

and implemented before initial receipt 

of byproduct, source, or SNM 

(excluding exempt quantities as 

described in § 30.18 of this chapter). 

(5) Security requirements must be 

implemented for the protection of SNM 

based on the type, enrichment, and 

quantity in accordance with 10 CFR part 

73, as applicable, and for the protection 

of Category 1 and Category 2 quantities 

of radioactive material in accordance 

with 10 CFR part 37, as applicable. 

(d) Fuel loading. (1)(i) An ML may 

authorize possession of a manufactured 

reactor into which the licensee has 

loaded fresh (unirradiated) fuel 

pursuant to a license issued under part 

70 of this chapter only if the 

manufactured reactor is configured 

during its loading, storage, and transport 

with at least two independent physical 

mechanisms in place, each of which is 

sufficient to prevent criticality assuming 

optimum neutron moderation and 

neutron reflection conditions. 

(ii) The ML applicant may file a 

separate, subsequent application for the 

10 CFR part 70 license or combine the 

application for the 10 CFR part 70 

license with the application for an ML. 

(iii) The Commission has determined 

that any such fueled manufactured 

reactor in which the independent 

physical mechanisms to prevent 

criticality have been installed is not in 

operation. 

(iv) Upon installation of the fueled 

manufactured reactor in its place of 

operation and a Commission finding 

that the acceptance criteria in the COL 

that authorized reactor construction are 

met under § 53.1452(g), the independent 

physical mechanisms to prevent 

criticality may be removed. Upon 

initiating the physical removal of any 

one of the independent physical 

mechanisms to prevent criticality, the 

fueled manufactured reactor has 

commenced operation. 

(2) Holders of 10 CFR part 70 licenses 

authorizing the possession and loading 

of fresh fuel into manufactured reactors 

must comply with the requirements of 

10 CFR part 70 for the facilities and 

activities related to the storage, 

movement, and loading of fresh fuel in 

the manufactured reactor. Holders of 

these 10 CFR part 70 licenses must 

comply with the requirements of 

Subpart H to 10 CFR part 70, regardless 

of whether their proposed activities 

meet the applicability criteria found in 

10 CFR 70.60. Procedures, equipment, 

and personnel required by the 10 CFR 

part 70 license, must be in place before 

the receipt of SNM at the manufacturing 

facility. 

(i) Before the receipt of SNM, the 

licensee must have security programs in 

place that meet the performance 

objectives of 10 CFR 73.67, with the 

following additions and exceptions: 

(A) A physical security plan 

describing the physical security 

program must be maintained and a 

cybersecurity program must be 

established for the possession and 

loading of fresh fuel into a 

manufactured reactor authorized by a 10 

CFR part 70 license, regardless of fuel 

type, enrichment, and quantity. 

(B) The physical security program 

must be designed to prevent unintended 

and uncontrolled criticality events. 

(C) The cybersecurity program must 

provide reasonable assurance that a 

cyberattack would not adversely impact 

the functions performed by digital assets 

used by the licensee for implementing 

the physical security requirements of 

this section, or the radiation monitoring 

and criticality requirements in this 

section or in 10 CFR part 70. 

(D) All holders of a part 70 license 

that authorizes loading of fresh fuel into 

a manufactured reactor must perform 

the screening required in § 73.67(d)(4) of 

this chapter to confirm the identity, 

trustworthiness, and reliability of 

individuals prior to granting unescorted 

access to special nuclear material; these 

determinations must be documented. 

(ii) [Reserved] 

(3) The loading or unloading of fresh 

fuel into or from a manufactured reactor 

and any changes to the configuration of 

reactivity control and prevention 

systems for the fueled manufactured 

reactor must be performed by a certified 

fuel handler meeting the requirements 

in subpart F of this part. 

(e) Transportation. (1) A holder of an 

ML may not transport or allow to be 

removed from the places of manufacture 

the manufactured reactor or portions 

thereof as defined in the ML except for 

transport to a site for which the 

Commission has issued a COL that 

references the subject ML. 

(2) A holder of an ML must include 

in any contract governing the transport 

of a manufactured reactor or portions 

thereof as defined in the ML from the 

places of manufacture to any other 

location, a provision requiring that the 

person transporting the manufactured 

reactor comply with all shipping 

requirements in applicable NRC 

regulations, certificates of compliance, 

and NRC-issued licenses. 

(3) Procedures governing the 

preparation of the manufactured reactor 

or portions thereof as defined in the ML 

for transport and the conduct of the 

transport must be issued prior to 

transport. The procedures must 

implement the protective measures and 

restrictions described in NRC 

regulations and NRC-issued licenses to 

protect the reactor from potential 

conditions that would adversely affect 

the safe operation of a commercial 

nuclear plant. 

(4) For a manufactured reactor that is 

to be loaded with fresh fuel before 

transport to the place of operation, the 

ML must specify that transportation will 

be in accordance with parts 71 and 73 

of this chapter. 

(f) Acceptance and installation at the 

site for which the Commission has 

issued a COL that references the subject 

ML. (1) Installation at the site for which 

the Commission has issued a COL that 

references the subject ML must follow 

the regulations in § 53.610. 

(2) Upon arrival at the site, the 

manufactured reactor or portions of a 

manufactured reactor may not be 

installed in its place of operation unless 

the COL holder performs inspections 

sufficient to verify the reactor is in 

compliance with the ML and has not 

been damaged in transit. The COL 

holder must perform these inspections 

in accordance with documented 

procedures subject to quality assurance 

measures commensurate with their 

importance to safety. In addition, 

inspections must confirm that the 

interface requirements between the 

manufactured reactor or portions of a 

manufactured reactor and the remaining 

portions of the commercial nuclear 

plant are met. 

Subpart F—Requirements for 

Operation 

§ 53.700 Operational objectives. 

(a) Each holder of an operating license 

(OL) or combined license (COL) under 

this part must develop, implement, and 

maintain controls for plant structures, 

systems, and components (SSCs), 

responsibilities of plant personnel, and 

plant programs during the operating life 

of each commercial nuclear plant such 

that the requirements defined in subpart 

B are satisfied. More specifically: 


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(1) Each holder of an OL or COL 

under this part must maintain the 

capabilities, availability, and reliability 

of plant SSCs to ensure that the safety 

functions identified in § 53.230 will be 

performed if called upon during 

licensing-basis events (LBEs). 

(2) Each holder of an OL or COL 

under this part must ensure that plant 

personnel have adequate knowledge and 

skills to perform their assigned duties 

that support the performance of the 

safety functions identified in § 53.230. 

(3) Each holder of an OL or COL 

under this part must implement plant 

programs sufficient to ensure that the 

safety functions identified in § 53.230 

will be performed if called upon during 

normal operations and LBEs. 

(b) [Reserved] 

§ 53.710 Maintaining capabilities and 

availability of structures, systems, and 

components. 

Controls must be provided for each 

commercial nuclear plant licensed 

under this part such that the 

capabilities, availability, and reliability 

of plant SSCs, when combined with 

corresponding programmatic controls 

and human actions, provide that the 

safety criteria defined in §§ 53.210 and 

53.220 will be met. 

(a) Technical specifications must be 

developed, implemented, and 

maintained that define conditions or 

limitations on plant operations that are 

necessary to ensure that safety-related 

(SR) SSCs can fulfill the safety functions 

identified under § 53.230 and support 

meeting the safety criteria of § 53.210. 

The technical specifications must 

describe the following requirements: 

(1) Limits on the inventory of 

radioactive materials within the reactor 

system and supporting systems with the 

potential, individually or collectively, to 

cause a release exceeding the safety 

criteria in § 53.210 as a result of a 

design-basis accident analyzed in 

accordance with § 53.450(f). 

(2) Operating limits for the facility 

that if exceeded could lead to a failure 

to perform a required safety function 

necessary to demonstrate compliance 

with the safety criteria in § 53.210. 

(3) For each SSC classified as SR in 

accordance with § 53.460, technical 

specifications must define— 

(i) Limiting conditions for operation. 

Limiting conditions for operation are 

the lowest functional capability or 

performance levels of SR SSCs required 

to ensure that the design-basis accidents 

analyzed in accordance with § 53.450(f) 

satisfy the safety criteria of § 53.210. 

When a limiting condition for operation 

is not met, the licensee must shut down 

the plant or follow any remedial action 

permitted by the technical 

specifications until the condition can be 

met. 

(ii) Surveillance requirements. 

Surveillance requirements are 

requirements relating to test, calibration, 

or inspection to assure that the 

necessary quality of systems and 

components is maintained and that the 

limiting conditions for operation will be 

met. 

(4) Design elements to be included are 

those elements of the plant such as 

materials of construction and geometric 

arrangements, which, if altered or 

modified, would have a significant 

effect on safety and are not covered in 

categories described in paragraphs (a)(1) 

through (3) of this section. 

(5) Administrative controls are the 

provisions relating to organization and 

management, procedures, 

recordkeeping, review and audit, and 

reporting necessary to assure operation 

of the plant in a safe manner. Each 

licensee must submit any reports to the 

Commission pursuant to approved 

technical specifications under § 53.040. 

(b) Controls on plant operations, 

including availability controls, must be 

developed and implemented to ensure 

that the configurations and special 

treatments for SR SSCs and non-safetyrelated but safety-significant (NSRSS) 

SSCs provide the capabilities, 

availability, and reliability required to 

demonstrate compliance with the 

criteria of §§ 53.220 and 53.450(e).1 The 

controls must— 

1The comprehensive risk metrics and 

related risk performance objectives 

established under § 53.220 involve assessing 

and averaging the risks over a defined period 

(e.g., plant year) and do not constitute a realtime requirement that must be continuously 

demonstrated by the licensee. 

(1)(i) Identify who within the 

commercial nuclear plant has authority 

to make configuration changes; 

(ii) Establish processes to make 

configuration changes to NSRSS SSCs; 

and 

(iii) Establish processes to ensure that 

all organizations of the commercial 

nuclear plant affected by the 

configuration changes are formally 

notified and approve of the change. 

(2) Describe how the special 

treatments for each NSRSS SSC and 

special treatments for SR SSCs beyond 

those under paragraph (a) of this section 

will be established and maintained over 

the operating life of the commercial 

nuclear plant. 

§ 53.715 Maintenance, repair, and 

inspection programs. 

(a) A program to control maintenance 

activities and monitor the performance 

or condition of SR and NSRSS SSCs 

must be developed, implemented, and 

maintained. 

(b) Whenever a licensee determines 

through activities related to 

maintenance, repair, and inspection of 

SSCs, the activities under § 53.710, or 

otherwise that the performance or 

condition of an SR or NSRSS SSC does 

not demonstrate compliance with 

established special treatments or 

performance goals related to 

capabilities, availability, or reliability, 

the licensee must take appropriate 

corrective action. 

(c) Performance and condition 

monitoring activities and associated 

goals and preventive maintenance 

activities must be evaluated at least 

every 24 months. The evaluations must 

take into account, where practical, 

industry-wide operating experience. 

Adjustments must be made where 

necessary to ensure that the objective of 

preventing failures of SSCs through 

maintenance is appropriately balanced 

against the objective of minimizing 

unavailability of SSCs due to 

monitoring or preventive maintenance. 

(d) Before performing maintenance 

activities (including but not limited to 

surveillance, post-maintenance testing, 

and corrective and preventive 

maintenance), the licensee must assess 

and manage the increase in risk that 

may result from the proposed 

maintenance activities. 

§ 53.720 Response to seismic events. 

If vibratory ground motion exceeding 

that of the operating basis earthquake 

ground motion or significant plant 

damage due to vibratory ground motion 

occurs, the licensee must shut down the 

commercial nuclear plant. If structures, 

systems, or components necessary for 

the safe shutdown of the commercial 

nuclear plant are not available after the 

occurrence of this vibratory ground 

motion, the licensee must consult with 

the Commission and must propose a 

plan for the timely, safe shutdown of the 

commercial nuclear plant. Prior to 

resuming operations, the licensee must 

demonstrate to the Commission that 

those features necessary for continued 

operation without undue risk to the 

health and safety of the public or 

necessary to maintain the licensing 

basis of the commercial nuclear plant 

were either not functionally damaged or 

have been repaired. 

§ 53.725 General staffing, training, 

personnel qualifications, and human factors 

requirements. 

(a) Two classes of commercial nuclear 

plants. Commercial nuclear plants 

licensed under this part are either of the 


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class, based upon the similarity of 

operating and technical characteristics 

of the plants in the class, of self-reliantmitigation facilities or of interactiondependent-mitigation facilities. A 

commercial nuclear plant is a selfreliant-mitigation facility if the U.S. 

Nuclear Regulatory Commission (NRC) 

determined as part of its approval of the 

OL or COL for that plant that its design 

demonstrates compliance with the 

criteria of § 53.800(a)(1) through (a)(5). 

Otherwise, the commercial nuclear 

plant is an interaction-dependentmitigation facility. 

(b) Purpose and applicability. The 

regulations in §§ 53.725 through 53.830 

address areas related to staffing, 

training, personnel qualifications, and 

human factors engineering for 

applicants for or holders of OLs or COLs 

under this part. These regulations are 

organized as follows: 

(1) Sections 53.725 through 53.745 

address general requirements for 

staffing, training, personnel 

qualifications, and human factors 

engineering. The regulations within 

these sections are applicable to all 

applicants for or holders of OLs or COLs 

under this part, except where 

specifically stated otherwise. 

(2) Sections 53.760 through 53.795 

address operator and senior operator 

licensing requirements. The regulations 

within these sections are applicable to 

those applicants for or holders of OLs or 

COLs under this part for interactiondependent-mitigation facilities that have 

not yet certified the permanent 

cessation of operations and permanent 

removal of fuel from the reactor vessel 

as described under § 53.1070. 

(3) Sections 53.800 through 53.820 

address generally licensed reactor 

operator requirements. The regulations 

within these sections are in lieu of 

§§ 53.760 through 53.795 for those 

applicants for or holders of OLs or COLs 

under this part for self-reliant-mitigation 

facilities that have not yet certified the 

permanent cessation of operations and 

permanent removal of fuel from the 

reactor vessel as described under 

§ 53.1070. 

(4) Section 53.830 provides general 

personnel training requirements. The 

regulations within this section are 

applicable to all applicants for or 

holders of OLs or COLs under this part. 

(c) Definitions. When used in 

§§ 53.725 through 53.830: 

Applicant refers to an applicant for an 

operator or senior operator license; 

licensee refers to the holder of an 

operator, senior operator, or generally 

licensed reactor operator license; and 

facility licensee refers to the licensee for 

the commercial nuclear plant where the 

applicant would be licensed or the 

licensee is licensed. 

Automation means a device or system 

that accomplishes (partially or fully) a 

function or task. 

Auxiliary operator means any 

individual who operates components of 

a commercial nuclear plant but does not 

manipulate controls or direct the 

manipulation of controls of the plant 

and is not required to be licensed under 

the provisions of this part. 

Controls when used with respect to a 

nuclear reactor means apparatus and 

mechanisms, the manipulation of which 

directly affects the reactivity or power 

level of the reactor. 

Generally licensed reactor operator 

means any individual licensed under 

the provisions of § 53.810 to manipulate 

controls of a self-reliant-mitigation 

facility and to direct the licensed 

activities of generally licensed reactor 

operators. 

Interaction-dependent-mitigation 

facility means a commercial nuclear 

plant design other than one that 

demonstrates compliance with the 

operating and technical characteristics 

defined under § 53.800. 

Load following means a commercial 

nuclear plant automatically changing its 

output to match expected demand in 

response to externally originated 

instructions or signals. 

Operator means any individual 

licensed under the provisions of 

§§ 53.760 through 53.795 to manipulate 

controls of an interaction-dependentmitigation facility. 

Performance testing means testing 

conducted to verify a simulation 

facility’s performance as compared to 

actual or predicted reference plant 

performance. 

Reference plant means the specific 

commercial nuclear plant on which a 

simulation facility’s configuration, 

system control arrangement, and design 

data are based. The reference plant may 

or may not be constructed. 

Self-reliant-mitigation facility means a 

commercial nuclear plant design that 

demonstrates compliance with the 

operating and technical characteristics 

defined under § 53.800. 

Senior operator means any individual 

licensed under the provisions of 

§§ 53.760 through 53.795 to manipulate 

controls of an interaction-dependentmitigation facility and to direct the 

licensed activities of operators. 

Simulation facility means an interface 

designed to provide a realistic imitation 

of the operation of a commercial nuclear 

plant used for the administration of 

examinations, for training, and/or to 

demonstrate compliance with 

experience requirements for applicants 

or licensees. A simulation facility may 

rely, in whole or part, upon the physical 

utilization of the reference plant itself. 

Systems approach to training means a 

training program that includes the 

following five elements: 

(1) Systematic analysis of the jobs to 

be performed. 

(2) Learning objectives derived from 

the analysis which describe desired 

performance after training. 

(3) Training design and 

implementation based on the learning 

objectives. 

(4) Evaluation of trainee mastery of 

the objectives during training. 

(5) Evaluation and revision of the 

training based on the performance of 

trained personnel in the job setting. 

§ 53.726 Communications. 

(a) An applicant or licensee or facility 

licensee must submit any 

communication or report required by 

the regulations contained within 

§§ 53.725 through 53.830 and must 

submit any application filed under these 

regulations to the Commission. 

(b) Each licensee that is required to 

comply with the requirements of 

§§ 53.760 through 53.795 (i.e., 

interaction-dependent-mitigation 

facilities) must notify the appropriate 

NRC contact within 30 days of the 

following in regard to a licensed 

operator or senior operator: 

(1) Permanent reassignment from the 

position for which the licensee has 

certified the need for a licensed operator 

or senior operator under § 53.775(a)(1); 

(2) Termination of any operator or 

senior operator; or 

(3) Permanent disability or illness as 

required under § 55.770 of this chapter. 

§ 53.728 Completeness and accuracy of 

information. 

Information provided to the 

Commission by an applicant for an 

operator or senior operator license or by 

a licensee or information required by 

statute or by the Commission’s 

regulations, orders, or license 

conditions to be maintained by the 

applicant or the licensee must be 

complete and accurate in all material 

respects. 

§ 53.730 Defining, fulfilling, and 

maintaining the role of personnel in 

ensuring safe operations. 

Each applicant for or holder of an OL 

or COL for a commercial nuclear plant 

under this part must comply with the 

following: 

(a) Human factors engineering design 

requirements. The plant design must 

reflect state-of-the-art human factors 

engineering principles for safe and 

reliable performance in all locations that 


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human activities are expected for 

performing or supporting the continued 

availability of plant safety or emergency 

response functions. 

(b) Human system interface design 

requirements. The plant design must 

provide for the following to support 

operating personnel in monitoring plant 

conditions and responding to plant 

events: 

(1) Features for displaying to 

operating personnel a minimum set of 

parameters that define the safety status 

of the plant and are capable of 

displaying both the full range of 

important plant parameters and data 

trends on demand, as well as indicating 

when process limits are being 

approached or exceeded; 

(2) Automatic indication of the 

bypassed and operable status of safety 

systems; 

(3) Direct indication of SSC status that 

relates to the ability of the SSC to 

perform its safety function, such as 

relief and safety valve position (i.e., 

open or closed) for barriers important to 

fulfilling safety functions of with such 

devices, and ultimate heat sink and 

cooling system status and availability; 

(4) Instrumentation to measure, 

record, and display key plant 

parameters related to the performance of 

SSCs and the integrity of barriers 

important to fulfilling safety functions 

to support operators in monitoring plant 

conditions and responding to plant 

events. Examples include temperatures 

and pressures within important systems 

or structures, core or fuel system 

conditions (including possible damage 

states), temperatures and levels 

associated with cooling functions, 

combustible gas concentrations, 

radiation levels in systems and within 

structures, and radioactive effluent 

releases; 

(5) Leakage control and detection in 

the design of systems that pass through 

barriers important to fulfilling safety 

functions for the release of 

radionuclides. An example is an SSC 

that penetrates a containment structure 

that might contain radioactive materials 

that could contribute to the source term 

during an accident; 

(6) Monitoring of in-plant radiation 

and airborne radioactivity as 

appropriate for a broad range of normal 

operating and accident conditions; and 

(7) For self-reliant-mitigation 

facilities, the plant design must also 

provide the generally licensed reactor 

operators with the capability to do the 

following: 

(i) Receive plant operating data, 

including reactor parameters and 

information needed for the evaluation of 

emergency conditions. 

(ii) Immediately initiate a reactor 

shutdown from their location. 

(iii) Promptly dispatch operations and 

maintenance personnel. 

(iv) Immediately implement 

responsibilities under the facility 

emergency plan, as applicable. 

(c) Concept of operations. A concept 

of operations that is of sufficient scope 

and detail to address the following must 

be provided: 

(1) Plant goals; 

(2) The roles and responsibilities of 

operating personnel and automation (or 

any combination thereof) that are 

responsible for completing plant 

functions; 

(3) Staffing, qualifications, and 

training; 

(4) The management of normal 

operations; 

(5) The management of off-normal 

conditions and emergencies; 

(6) The management of maintenance 

and modifications; and 

(7) The management of tests, 

inspections, and surveillances. 

(d) Functional requirements analysis 

and function allocation. A functional 

requirements analysis and a function 

allocation must be provided that are 

sufficient to demonstrate compliance 

with the following: 

(1) The functional requirements 

analysis must address how safety 

functions and functional safety criteria 

are satisfied, and 

(2) The function allocation must 

describe how the safety functions will 

be assigned to human action, 

automation, active safety features, 

passive safety features, and/or inherent 

safety characteristics. 

(e) Operating experience. A program, 

during construction and during 

operation, as applicable, for evaluating 

and applying operating experience must 

be developed, implemented, and 

maintained. 

(f) Staffing plan. A staffing plan must 

be developed and comply with the 

following: 

(1) The staffing plan must include a 

description of how engineering 

expertise will be available to the onshift operating personnel during all 

plant conditions, to assist if they 

encounter a situation not covered by 

procedures or training. Engineering 

expertise includes familiarity with the 

operation of the plant for which the 

expertise is provided and one of the 

following: 

(i) A bachelor’s degree in engineering, 

engineering technology, or physical 

science from an institution accredited 

by a U.S. government recognized 

accrediting body or equivalent; or 

(ii) A Professional Engineer’s license 

from a U.S. State or territory. 

(2) Applicants for or holders of OLs or 

COLs for interaction-dependentmitigation facilities must include within 

their staffing plans a description of how 

the proposed numbers, positions, and 

qualifications of operators and senior 

operators across all modes of plant 

operations will be sufficient to ensure 

that plant safety functions will be 

maintained. This description must be 

supported by human factors engineering 

analyses and assessments. 

(3) Applicants for or holders of OLs or 

COLs for self-reliant-mitigation facilities 

must include within their staffing plans 

a description of how generally licensed 

reactor operator staffing that is both 

sufficient to continually monitor the 

operations of fueled reactors and to 

provide for a continuity of 

responsibility for facility operations at 

all times during the operating phase will 

be maintained. 

(4) Applicants for or holders of OLs or 

COLs under this part must include 

within their staffing plans a description 

of how the numbers, positions, and 

responsibilities of personnel contained 

within those plans will adequately 

support all necessary functions within 

areas such as plant operations, 

equipment surveillance and 

maintenance, radiological protection, 

chemistry control, fire brigades, 

engineering, security, and emergency 

response. 

(5) The staffing plan must be 

approved by the NRC as part of its 

approval of the OL or COL for the plant. 

The approved staffing plan is subject to 

the requirements of § 53.1565. 

(g) Training, examination, and 

proficiency programs. Develop, 

implement, and maintain programs that 

comply with the following 

requirements. These programs must be 

approved by the NRC as part of its 

approval of the OL or COL for the plant: 

(1) For those applicants for or holders 

of OLs or COLs for interactiondependent-mitigation facilities: 

(i) The operator licensing initial 

training program required under 

§ 53.780(a); 

(ii) The operator licensing initial 

examination program required under 

§ 53.780(b); 

(iii) The operator licensing 

requalification program required under 

§ 53.780(c); and 

(iv) The operator proficiency program 

required under § 53.780(g). 

(2) For those applicants for or holders 

of OLs or COLs for self-reliantmitigation facilities, the generally 

licensed reactor operator training, 

examination, and proficiency programs 

required under § 53.815. 


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(3) The operator licensing 

requalification programs required under 

§ 53.780(c) or § 53.815(b) must be 

implemented upon commencing the 

administration of initial examinations 

under the operator licensing 

examination program required under 

§ 53.780(b) or § 53.815(b), respectively. 

§ 53.735 General exemptions. 

The regulations in §§ 53.725 through 

53.830 do not require a license for an 

individual who— 

(a) Under the direction and in the 

presence of an operator or senior 

operator or a generally licensed reactor 

operator, as appropriate, manipulates 

the controls of a commercial nuclear 

plant as a part of the individual’s 

training in a facility licensee’s training 

program as approved by the 

Commission to qualify for an operator or 

senior operator license or a generally 

licensed reactor operator license there, 

as appropriate, under these regulations; 

or 

(b) Under the direction and in the 

presence of a senior operator or 

generally licensed reactor operator, as 

appropriate, manipulates the controls of 

a commercial nuclear plant to load or 

unload the fuel into, out of, or within 

the reactor vessel while the reactor is 

not operating. 

§ 53.740 Facility licensee requirements— 

General. 

(a) Facility licensees must 

demonstrate compliance with the 

requirements of either §§ 53.760 through 

53.795 for interaction-dependentmitigation facilities or §§ 53.800 through 

53.820 for self-reliant-mitigation 

facilities. 

(b) The facility licensee must 

maintain the staffing complement 

described under its approved facility 

staffing plan until such time as the 

permanent cessation of operations and 

permanent removal of fuel from the 

reactor vessel has been certified as 

described under § 53.1070. The 

approved staffing plan is subject to the 

requirements of § 53.1565. 

(c) Except as provided under § 53.735, 

the facility licensee may not permit the 

manipulation of the controls of a 

commercial nuclear plant by anyone 

who is not an operator or senior 

operator or generally licensed reactor 

operator, as appropriate. 

(d) Facility licensees for interactiondependent-mitigation facilities that have 

not yet certified the permanent 

cessation of operations and permanent 

removal of fuel from the reactor vessel 

as described under § 53.1070 must 

designate senior operators to be 

responsible for supervising the licensed 

activities of operators. 

(e) Apparatus and mechanisms other 

than controls, the operation of which 

may affect the reactivity or power level 

of a reactor, must be manipulated only 

while plant conditions are being 

monitored by an individual who is an 

operator or senior operator or a 

generally licensed reactor operator, as 

appropriate. 

(f)(1) Load following is permitted if at 

least one of the following is 

immediately capable of refusing 

demands when they could challenge the 

safe operation of the plant or when 

precluded by the plant equipment 

conditions: 

(i) The actuation of an automatic 

protection system that utilizes setpoints 

more conservative than those otherwise 

credited for the purposes of reactor 

protection; or 

(ii) An automated control system; or 

(iii) An operator or senior operator or 

a generally licensed reactor operator, as 

appropriate. 

(2) The provisions of paragraph (e) of 

this section do not apply during load 

following operations. 

(g)(1) Facility licensees for 

interaction-dependent-mitigation 

facilities must have present during 

alteration of the core (including fuel 

loading or transfer) an individual 

holding a senior operator license, or a 

senior operator license limited to fuel 

handling to directly supervise the 

activity and, during this time, the 

facility licensee must not assign other 

duties to this person. 

(2) Facility licensees for self-reliantmitigation facilities must have present 

during alteration of the core (including 

fuel loading or transfer) an individual 

holding a generally licensed reactor 

operator license to directly supervise 

the activity and, during this time, the 

facility licensee must not assign other 

duties to this person. 

(3) The provisions of paragraphs (g)(1) 

and (2) of this section do not apply to 

core alterations performed as part of 

refueling operations while a facility that 

is capable of online refueling is 

operating at power. 

(h) Facility licensees may take 

reasonable action that departs from a 

license condition or a technical 

specification (contained in a license 

issued under this part) in an emergency 

when this action is immediately needed 

to protect the public health and safety 

and no action consistent with license 

conditions and technical specifications 

that can provide adequate or equivalent 

protection is immediately apparent. 

Such facility licensee action must be 

approved, as a minimum, by a senior 

operator or a generally licensed reactor 

operator, as applicable, or, after 

certifying the permanent cessation of 

operations and permanent removal of 

fuel from the reactor vessel as described 

under § 53.1070 by a certified fuel 

handler, senior operator, or generally 

licensed reactor operator, as applicable, 

prior to taking the action. 

§ 53.745 Operator license requirements. 

A person must be authorized by a 

license issued by the Commission to 

perform the function of an operator, 

senior operator, or generally licensed 

reactor operator as defined in this part. 

§ 53.760 Operator licensing. 

(a) Applicability. Sections 53.760 

through 53.795 address operator and 

senior operator licensing requirements. 

The regulations within these sections 

are applicable to those applicants for or 

holders of OLs or COLs under this part 

for interaction-dependent-mitigation 

facilities that have not yet certified the 

permanent cessation of operations and 

permanent removal of fuel from the 

reactor vessel as described under 

§ 53.1070. 

(b) Reserved. 

§ 53.765 Medical requirements. 

(a) An applicant for an operator or 

senior operator license must have a 

medical examination by a physician. An 

operator or senior operator must have a 

medical examination by a physician 

every 2 years. 

(b) To certify the medical fitness of an 

applicant for an operator or senior 

operator license, an authorized 

representative of the facility licensee 

must complete and sign NRC Form 396, 

‘‘Certification of Medical Examination 

by Facility Licensee,’’ which can be 

obtained by writing the Office of the 

Chief Information Officer, U.S. Nuclear 

Regulatory Commission, Washington, 

DC 20555–0001, by calling 301–415– 

7232, or by visiting the NRC’s website 

at https://www.nrc.gov and selecting 

forms from the index found on the home 

page, or by other means provided by the 

NRC. 

(1) Form NRC 396 must certify that a 

physician has conducted the medical 

examination of the applicant as required 

in paragraph (a) of this section. 

(2) When the medical certification 

requests a conditional license based on 

medical evidence, the medical evidence 

must be submitted on NRC Form 396 to 

the Commission to enable the 

Commission to make a determination in 

accordance with § 53.775(b). 

(c) The facility licensee must 

document and maintain the results of 

medical qualifications data, test results, 


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and each operator’s or senior operator’s 

medical history for the current license 

period and provide the documentation 

to the Commission upon request. The 

facility licensee must retain this 

documentation while an individual 

performs the functions of an operator or 

senior operator. 

§ 53.770 Incapacitation because of 

disability or illness. 

If, during the term of the operator or 

senior operator license, the licensee 

develops a permanent physical or 

mental condition that causes the 

licensee to fail to demonstrate 

compliance with the requirements of 

§ 53.775(b)(1)(i), the facility licensee 

must notify the Commission within 30 

days of learning of the diagnosis. For 

conditions for which a conditional 

license (as described in § 53.775(b)) is 

requested, the facility licensee must 

provide medical certification on Form 

NRC 396 to the Commission (as 

described in § 53.765(b)). 

§ 53.775 Applications for operators and 

senior operators. 

(a) How to apply. (1) The applicant for 

an operator or senior operator license 

must— 

(i) Complete NRC Form 398, 

‘‘Personal Qualification Statement— 

Licensee,’’ which can be obtained by 

writing the Office of the Chief 

Information Officer, U.S. Nuclear 

Regulatory Commission, Washington, 

DC 20555–0001, by calling 301–415– 

5877, or by visiting the NRC’s website 

at https://www.nrc.gov and selecting 

forms from the index found on the home 

page, or by other means provided by the 

NRC; 

(ii) File an original of NRC Form 398, 

or an equivalent electronic submittal, 

together with the information required 

in paragraphs (a)(1)(iii) and (a)(1)(iv) of 

this section, with the appropriate 

Regional Administrator. 

(iii) Provide evidence that the 

applicant, as a trainee, has successfully 

demonstrated competence in 

manipulating the controls of either the 

facility for which a license is sought or 

a simulation facility that demonstrates 

compliance with the requirements of 

§ 53.780(e). For operators applying for a 

senior operator license, certification that 

the operator has successfully operated 

the controls of the facility as an operator 

will be accepted; and 

(iv) Provide certification by the 

facility licensee of medical condition 

and general health on Form NRC 396, to 

comply with § 53.765. 

(2) The Commission may at any time 

after the application has been filed, and 

before the license has expired, require 

further information under oath or 

affirmation to enable it to determine 

whether to grant or deny the application 

or whether to revoke, modify, or 

suspend the license. 

(3) An applicant whose application 

has been denied because of a medical 

condition or their general health may 

submit a further medical report at any 

time as a supplement to the application. 

(4) Each application and statement 

must contain complete and accurate 

disclosure as to all matters required to 

be disclosed. The applicant must sign 

statements required by paragraphs 

(a)(1)(i) and (a)(1)(ii) of this section. 

(b) Disposition of an initial 

application. (1) License approval. The 

Commission will approve an initial 

application if it finds that the following 

criteria are met: 

(i) Health. The applicant’s medical 

condition and general health will not 

adversely affect the performance of 

assigned operator or senior operator job 

duties or cause operational errors 

endangering public health and safety. 

The Commission will base its finding 

upon the certification by the facility 

licensee as detailed in § 53.765(b). 

(ii) Examination. The applicant has 

passed the requisite examination in 

accordance with § 53.780(b). The 

examination determines whether the 

applicant for an operator’s or senior 

operator’s license has learned to operate 

a facility competently and safely, and 

additionally, in the case of a senior 

operator, whether the applicant has 

learned to supervise the licensed 

activities of operators competently and 

safely. 

(2) Conditional license. If an 

applicant’s general medical condition 

does not demonstrate compliance with 

the minimum standards under 

§ 53.775(b)(1)(i) of this section, the 

Commission may approve the 

application and include conditions in 

the license to accommodate the medical 

condition. The Commission will 

consider the recommendations and 

supporting evidence of the facility 

licensee and of the examining physician 

(provided on Form NRC 396) in arriving 

at its decision. 

(c) Re-applications. (1) An applicant 

whose application for a license has been 

denied because of failure to pass the 

examination may file a new application. 

The application must be submitted on 

Form NRC 398 and include a statement 

signed by an authorized representative 

of the facility licensee by whom the 

applicant will be employed that states 

in detail the extent of the applicant’s 

additional training and remediation 

since the denial and certifies that the 

applicant is ready for re-examination. 

(2) An applicant who has passed a 

portion of the examination and failed 

another may request in a new 

application on Form NRC 398 to be 

excused from re-examination on the 

portions of the examination that the 

applicant has passed. The Commission 

may in its discretion grant the request 

if it determines that sufficient 

justification is presented. 

§ 53.780 Training, examination, and 

proficiency program. 

(a) Operator licensing initial training 

program. (1) A program that is based 

upon a systems approach to training, as 

defined by § 53.725(b), must be utilized 

for the training of applicants for 

operator and senior operator licenses. 

The program must ensure that 

applicants at the facility will possess the 

knowledge, skills, and abilities 

necessary to protect the public health 

and maintain those plant safety 

functions specific to the facility design. 

The program must be approved by the 

Commission prior to its use for training 

applicants, as described under 

§ 53.730(g). The approved operator 

licensing initial training program is 

subject to the requirements of § 53.1565. 

(2) The facility licensee must 

maintain operator licensing initial 

training program records documenting 

the initial operator licensing training 

administered and completed by each 

applicant. The facility licensee must 

retain these records during the period in 

which any trainees subsequently remain 

licensed as operators or senior operators 

at the facility. 

(b) Operator licensing initial 

examination program. (1) The facility 

licensee must establish and implement 

an examination program for testing a 

representative sample of the knowledge, 

skills, and abilities needed to safely 

perform operator and senior operator 

duties, to include both the examination 

methods and criteria to be used to assess 

passing performance. The program must 

provide for valid and reliable 

examinations and be approved by the 

Commission prior to its use for 

examining applicants, as described 

under § 53.730(g). The approved 

operator licensing initial examination 

program is subject to the requirements 

of § 53.1565. 

(2) The facility licensee must submit 

prepared examinations to the 

Commission for review and approval in 

advance of their administration. 

(3) The Commission will either 

administer an approved examination or 

allow the facility licensee to administer 

the examination. The facility licensee 

must ensure that sufficient advance 

notification is provided to the 


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Commission to either administer the 

examination or allow for a 

representative of the Commission to be 

afforded the opportunity to be present 

when the facility licensee administers 

the examination. 

(4) Graded examination 

documentation for each applicant must 

be promptly provided to the 

Commission for review in making 

operator licensing decisions. 

(5) The facility licensee must 

maintain operator licensing initial 

examination program records 

documenting the participation of each 

operator and senior operator applicant 

in the initial examination. The records 

must contain copies of examinations 

administered, the answers given by the 

applicant, documentation of the grading 

of examinations, and documentation of 

any additional training administered in 

areas in which an applicant exhibited 

deficiencies. The facility licensee must 

retain these records during the period in 

which the associated operators or senior 

operators remain licensed at the facility. 

(c) Operator licensing requalification 

program. (1) A program based upon a 

systems approach to training, as defined 

by § 53.725(b), must be utilized for the 

continuing training of operators and 

senior operators. 

(i) The program must ensure that 

operators and senior operators at the 

facility maintain the knowledge, skills, 

and abilities necessary to protect the 

public health and maintain those plant 

safety functions specific to the facility 

design. The program must be conducted 

for a continuous period not to exceed 24 

months in duration. 

(ii) The program must be approved by 

the Commission prior to its use for 

continuing training, as described under 

§ 53.730(g). The approved operator 

licensing requalification program is 

subject to the requirements of § 53.1565. 

(2) The following requirements apply 

to operator licensing requalification 

programs: 

(i) The facility licensee must propose 

a requalification examination program 

for testing, for each requalification 

period, a sample of the topics included 

under the systems approach to training, 

to include both the examination 

methods and criteria to be used to assess 

passing performance. The program must 

provide for valid and reliable 

examinations and be approved by the 

Commission prior to its use for 

examining operators and senior 

operators, as described under 

§ 53.730(g). The approved 

requalification examination program is 

subject to the requirements of § 53.1565. 

(ii) The following requirements apply 

to the requalification examination 

program: 

(A) The facility licensee must make 

prepared requalification examinations 

available to the Commission for review. 

(B) The facility licensee must ensure 

that a representative of the Commission 

is afforded the opportunity to be present 

during requalification examination 

administration. 

(C) The facility licensee must ensure 

that each operator and senior operator is 

administered a complete requalification 

examination on a periodicity not to 

exceed 24 months. Additionally, the 

facility licensee must ensure that any 

licensed operator or senior licensed 

operator who either demonstrates 

unsatisfactory performance on, or fails 

to complete, the biennial requalification 

examination is removed from the 

performance of licensed operator and 

senior licensed operator duties until 

such time that any necessary remedial 

training has been completed and a 

retake examination has been passed. 

(D) The facility licensee must 

promptly provide a summary of 

examination results for each operator 

and senior operator following the 

completion of the requalification 

examination. 

(3) The facility licensee must 

maintain operator licensing 

requalification program records 

documenting the participation of each 

operator and senior operator in the 

requalification program. The records 

must contain copies of examinations 

administered, the answers given by the 

operator or senior operator, 

documentation of the grading of 

examinations, and documentation of 

any additional training administered in 

areas in which an operator or senior 

operator exhibited deficiencies. The 

facility licensee must retain these 

records until the operator’s or senior 

operator’s license is renewed. 

(d) Examination integrity. Applicants, 

operators and senior operators, and 

facility licensees must not engage in any 

activity that compromises the integrity 

of any application or examination 

required by §§ 53.760 through 53.795. 

The integrity of an examination is 

considered compromised if any activity, 

regardless of intent, affected, or, but for 

detection, could have affected the 

equitable and consistent administration 

of the examination. This includes 

activities related to the preparation and 

certification of applications and all 

activities related to the preparation, 

administration, and grading of 

examinations required by §§ 53.760 

through 53.795. 

(e) Simulation facilities. (1) This 

section addresses the use of a 

simulation facility for the 

administration of examinations, for 

training, or to demonstrate compliance 

with experience requirements for 

applicants for operator and senior 

operator licenses. 

(2) Simulation facilities used for 

training purposes, for demonstrating 

compliance with experience 

requirements, or for the conduct of 

examinations under § 53.780(b) and (c) 

must demonstrate compliance with the 

following criteria as they relate to the 

facility licensee’s reference plant: 

(i) The simulation facility must be of 

sufficient scope and fidelity for 

individuals to acquire and demonstrate 

the necessary knowledge, skills, and 

abilities to safely perform operator and 

senior operator duties. 

(ii) The simulation facility must 

utilize models relating to nuclear, 

thermal-hydraulic, and other applicable 

design-specific characteristics that 

either replicate the most recent fuel load 

in the reference commercial nuclear 

plant or, prior to initial fuel load (or, for 

a fueled manufactured reactor, prior to 

initiating the physical removal of any 

one of the independent physical 

mechanisms to prevent criticality 

required under § 53.620(d)(1)), replicate 

the intended initial fuel load for the 

reference commercial nuclear plant, 

with the exception of those portions of 

the simulation facility that utilize the 

reference plant itself. 

(iii) Simulation facility fidelity must 

be demonstrated so that significant 

control manipulations are completed 

without procedural exceptions, 

simulator performance exceptions, or 

deviation from the approved training 

scenario sequence. 

(3) Facility licensees that maintain a 

simulation facility that has been 

approved by the Commission for 

training purposes, demonstrating 

compliance with experience 

requirements, or the conduct of 

examinations under § 53.780(b) and (c) 

for the facility licensee’s reference plant 

must: 

(i) Conduct performance testing 

throughout the life of the simulation 

facility in a manner sufficient to ensure 

that paragraph (e)(2) of this section is 

met; 

(ii) Retain the results of performance 

testing for 4 years after the completion 

of each performance test or until 

superseded by updated test results; 

(iii) Promptly correct modeling and 

hardware discrepancies and 

discrepancies identified from scenario 

validation and from performance testing 

or provide justification as to why the 


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presence of such discrepancies will not 

adversely affect simulator performance 

with respect to the criteria of paragraph 

(e)(2) of this section; 

(iv) Make the results of any 

uncorrected performance test failures 

that may exist at the time of the initial 

license examination or requalification 

examination available for NRC review, 

prior to or concurrent with preparations 

for each initial license examination or 

requalification examination; and 

(v) Maintain the provisions for license 

application and examination integrity 

consistent with § 53.780(d). 

(4) A simulation facility must 

demonstrate compliance with the 

requirements of paragraphs (e)(2) and 

(e)(3) of this section for the Commission 

to accept the simulation facility for 

conducting initial examinations as 

described in § 53.780(b), requalification 

training as described in § 53.780(c), or 

performing control manipulations that 

affect reactivity to establish eligibility 

for an operator or senior operator 

license as described in § 53.775(a). 

(f) Waiver of examination 

requirement. On application, the 

Commission may waive any or all of the 

requirements for an examination if it 

finds that the applicant has 

demonstrated the required knowledge, 

skills, and abilities to safely operate the 

plant, and is capable of continuing to do 

so. The Commission may make such a 

finding based on demonstration of the 

following: 

(1) Operating experience at a 

comparable facility; 

(2) Proof of the applicant’s past 

competent and safe performance; and 

(3) Proof of the applicant’s current 

qualifications. 

(g) Proficiency. The facility licensee 

must develop, implement, and maintain 

a proficiency program to ensure that 

operators and senior operators will 

actively perform the functions of an 

operator or senior operator, respectively, 

as needed to maintain proficiency with 

on-shift duties and familiarity with 

plant status. This program must include 

those steps that will be taken to reestablish proficiency when it cannot be 

maintained. This program must be 

approved by the Commission as part of 

its approval of the OL or COL for the 

plant. The approved proficiency 

program is subject to the requirements 

of § 53.1565. 

(h) Records. Each record required by 

this section must be legible throughout 

the retention period specified by each 

Commission regulation. The record may 

be the original, a reproduced copy, or an 

electronic copy provided that the copy 

is authenticated by authorized 

personnel. 

§ 53.785 Conditions of operator and senior 

operator licenses. 

Each operator and senior operator 

license contains and is subject to the 

following conditions whether stated in 

the license or not: 

(a) Neither the license nor any right 

under the license may be assigned or 

otherwise transferred. 

(b) The license is limited to the 

facility for which it is issued. 

(c) The license is limited to those 

controls of the facility or facilities 

specified in the license. 

(d) The license is subject to, and the 

licensee must observe, all applicable 

rules, regulations, and orders of the 

Commission. 

(e) The licensee must maintain or reestablish proficiency in accordance with 

the facility licensee’s Commissionapproved proficiency program required 

under § 53.780(g). 

(f) The licensee must be subject to the 

facility’s Commission-approved 

operator licensing requalification and 

requalification examination programs 

required under § 53.780(c). 

(g) The licensee must have a biennial 

medical examination as described by 

§ 53.765. 

(h) The licensee must notify the 

Commission within 30 days about a 

conviction for a felony. 

(i) The licensee must not consume or 

ingest alcoholic beverages within the 

protected area of commercial nuclear 

plants. The licensee must not use, 

possess, or sell any illegal drugs. The 

licensee must not perform activities 

authorized by a license issued under 

this part while under the influence of 

alcohol or any prescription, over-thecounter, or illegal substance that could 

adversely affect his or her ability to 

safely and competently perform his or 

her licensed duties. For the purpose of 

this paragraph, with respect to alcoholic 

beverages and drugs, the term ‘‘under 

the influence’’ means the licensee 

exceeded, as evidenced by a confirmed 

test result, the lower of the cutoff levels 

for drugs or alcohol contained in 10 CFR 

part 26, or as established by the facility 

licensee. The term ‘‘under the 

influence’’ also means the licensee 

could be mentally or physically 

impaired as a result of substance use 

including prescription and over-thecounter drugs, as determined under the 

provisions, policies, and procedures 

established by the facility licensee for 

its fitness-for-duty program, in such a 

manner as to adversely affect his or her 

ability to safely and competently 

perform licensed duties. 

(j) Each licensee must participate in 

the drug and alcohol testing programs as 

required under 10 CFR part 26. 

(k) The licensee must comply with 

any other conditions that the 

Commission may impose to protect 

health or to minimize danger to life or 

property. 

§ 53.790 Issuance, modification, and 

revocation of operator and senior operator 

licenses. 

(a) Issuance of operator and senior 

operator licenses. If the Commission 

determines that an applicant for an 

operator license or a senior operator 

license demonstrates compliance with 

the requirements of the Atomic Energy 

Act of 1954, as amended, (the Act) and 

its regulations, it will issue a license in 

the form and containing any conditions 

and limitations it considers appropriate 

and necessary. 

(b) Modification and revocation of 

operator and senior operator licenses. 

(1) The terms and conditions of all 

operator and senior operator licenses are 

subject to amendment, revision, or 

modification by reason of rules, 

regulations, or orders issued in 

accordance with the Act or any 

amendments thereto. 

(2) Any license may be revoked, 

suspended, or modified, in whole or in 

part— 

(i) For any material false statement in 

the application or in any statement of 

fact required under section 182 of the 

Act; 

(ii) Because of conditions revealed by 

the application or statement of fact or 

any report, record, inspection, or other 

means that would warrant the 

Commission to refuse to grant a license 

on an original application; 

(iii) For willful violation of, or failure 

to observe, any of the terms and 

conditions of the Act or the license, or 

of any rule, regulation, or order of the 

Commission; 

(iv) For any conduct determined by 

the Commission to be a hazard to safe 

operation of the facility; or 

(v) For the sale, use, or possession of 

illegal drugs, or refusal to participate in 

the facility drug and alcohol testing 

program, or a confirmed positive test for 

drugs, drug metabolites, or alcohol in 

violation of the conditions and cutoff 

levels established by § 53.785(i) or the 

consumption of alcoholic beverages 

within the protected area of commercial 

nuclear plants, or a determination of 

unfitness for scheduled work as a result 

of the consumption of alcoholic 

beverages. 

§ 53.795 Expiration and renewal of 

operator and senior operator licenses. 

(a) Expiration. (1) Each operator 

license and senior operator license 

expires 6 years after the date of 


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issuance, upon termination of 

employment with the facility licensee, 

or upon determination by the facility 

licensee that the licensed individual no 

longer needs to maintain a license. 

(2) If a licensee files an application for 

renewal or an upgrade of an existing 

license on Form NRC 398 at least 30 

days before the expiration of the 

existing license, it does not expire until 

disposition of the application for 

renewal or for an upgraded license has 

been finally determined by the 

Commission. Filing by mail will be 

deemed to be complete at the time the 

application is postmarked 

(b) Renewal. (1) The applicant for 

renewal of an operator license or senior 

operator license must— 

(i) Complete and sign Form NRC 398 

and include the number of the license 

for which renewal is sought. 

(ii) File an original of NRC Form 398 

as specified in § 53.775. 

(iii) Provide written evidence of the 

applicant’s experience under the 

existing license and the approximate 

number of hours that the licensee has 

operated the facility. 

(iv) Provide a statement by an 

authorized representative of the facility 

licensee that during the effective term of 

the current license the applicant has 

satisfactorily completed the 

requalification program for the facility 

for which operator or senior operator 

license renewal is sought. 

(v) Provide evidence that the 

applicant has discharged the license 

responsibilities competently and safely. 

The Commission may accept as 

evidence of the applicant’s having met 

this requirement a certificate of an 

authorized representative of the facility 

licensee or holder of an authorization by 

which the licensee has been employed. 

(vi) Provide certification by the 

facility licensee of medical condition 

and general health on Form NRC 396, to 

comply with § 53.765. 

(2) The license will be renewed if the 

Commission finds that— 

(i) The medical condition and the 

general health of the licensee continue 

to be such as not to cause operational 

errors that endanger public health and 

safety. The Commission will base this 

finding upon the certification by the 

facility licensee as described in 

§ 53.765(b). 

(ii) The licensee— 

(A) Is capable of continuing to 

competently and safely assume licensed 

duties; 

(B) Has successfully completed a 

requalification program that has been 

approved by the Commission as 

required by § 53.780(c); and 

(C) Has passed the requalification 

examinations as required by § 53.780(c). 

(iii) There is a continued need for an 

operator to operate or for a senior 

operator to supervise operators at the 

facility designated in the application. 

(iv) The past performance of the 

licensee has been satisfactory to the 

Commission. In making its finding, the 

Commission will include in its 

evaluation information such as notices 

of violations or letters of reprimand in 

the licensee’s docket. 

§ 53.800 Facility licensees for self-reliantmitigation facilities. 

(a) A commercial nuclear plant is a 

self-reliant-mitigation facility if the NRC 

determined as part of its approval of the 

OL or COL for that plant that its design 

demonstrates compliance with criteria 

(a)(1) though (a)(5) of this section. A 

self-reliant-mitigation facility is of a 

class, based upon the similarity of 

operating and technical characteristics 

of the plants in the class, such that its 

licensee must comply with the 

requirements of §§ 53.800 through 

53.820 in lieu of those in §§ 53.760 

through 53.795. 

(1) The safety performance criteria of 

§§ 53.210 and 53.220 and, if applicable, 

any alternative criteria used in 

accordance with § 53.470, must be met 

without reliance upon human action for 

credited event mitigation. 

(2) The results of a probabilistic risk 

analysis must demonstrate that the 

evaluation criteria for the events 

analyzed in accordance with § 53.450 

will be met without reliance on human 

actions to achieve acceptable event 

mitigation. 

(3) The functional requirements 

analysis and function allocation 

performed under § 53.730(d) must 

demonstrate that functions required for 

safety are not reliant upon credited 

human action. 

(4) The plant response to events 

analyzed under § 53.450 must rely 

exclusively on safety features and 

characteristics that will neither be 

rendered unavailable by credible human 

errors of commission or omission nor 

credibly require manual human 

operation in response to equipment 

failures. Compliance with this 

paragraph may be achieved through the 

use of SSCs that function through 

inherent characteristics or that have 

engineered protections against human 

failures. 

(5) The plant design must provide for 

a layered defense-in-depth approach 

that is not dependent upon any single 

barrier or credited human action. 

(b) [Reserved] 

§ 53.805 Facility licensee requirements 

related to generally licensed reactor 

operators. 

(a) Licensees for self-reliantmitigation facilities that have not yet 

certified the permanent cessation of 

operations and permanent removal of 

fuel from the reactor vessel as described 

under § 53.1070 must demonstrate 

compliance with the following 

requirements: 

(1) Ensure that, in addition to being 

qualified to perform those items 

identified by the facility-specific 

systems approach to training conducted 

under § 53.815, generally licensed 

reactor operators are qualified to safely 

and competently— 

(i) Perform administrative tasks, 

including compliance with technical 

specifications, and perform operability 

determinations; 

(ii) Implement maintenance and 

configuration controls; 

(iii) Comply with radioactive release 

limitations; 

(iv) Understand plant operating data, 

including reactor parameters, and 

evaluate emergency conditions; 

(v) Initiate a reactor shutdown from 

necessary locations; 

(vi) Dispatch and direct operations 

and maintenance personnel; 

(vii) Implement any applicable 

responsibilities under the facility 

emergency plan; and 

(viii) Make required notifications to 

local, State, participating Tribal and 

Federal authorities. 

(2) Develop, implement, and maintain 

facility technical specifications that 

provide the necessary administrative 

controls to ensure the implementation 

of these requirements. 

(3) Develop, implement, and maintain 

the generally licensed reactor operator 

training, examination, and proficiency 

programs required under § 53.815. 

(4) Ensure that generally licensed 

reactor operators are subject to the 

facility’s generally licensed reactor 

operator training, examination, and 

proficiency programs required under 

§ 53.815. Ensure that generally licensed 

reactor operators are subject to and 

comply with the applicable 

programmatic requirements for plant 

personnel required under 10 CFR parts 

26 and 73. An individual that is not in 

compliance with any of these programs 

is not qualified to be in a position that 

may involve the manipulation of the 

controls of the commercial nuclear 

plant. 

(5) Report annually to the NRC the 

identity of all generally licensed reactor 

operators at the commercial nuclear 

plant, including all additions and 

deletions since the previous report. 


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(6) Ensure that the facility design 

continues to meet the criteria of 

§ 53.800. 

(b) [Reserved] 

§ 53.810 Generally licensed reactor 

operators. 

(a) A general license to manipulate 

the controls of a self-reliant-mitigation 

facility and to direct the licensed 

activities of generally licensed reactor 

operators is hereby issued to any 

individual employed in a position that 

may involve the manipulation of the 

controls of that self-reliant-mitigation 

facility and who observes the 

restrictions of this section. 

(b) A generally licensed reactor 

operator must comply with the 

operating procedures and other 

conditions specified in the license 

authorizing operation of the facility. 

(c) The general license is limited to 

the facility or facilities at which the 

operator is employed. 

(d) The Commission will suspend the 

general license on an individual basis 

for violations of any provision of the Act 

or any rule or regulation issued 

thereunder whenever the Commission 

deems such suspension desirable, 

including— 

(1) For willful violation of, or failure 

to observe, any of the terms and 

conditions of the Act or the general 

license, or of any rule, regulation, or 

order of the Commission; 

(2) For any conduct determined by the 

Commission to be a hazard to safe 

operation of the facility; or 

(3) For the sale, use, or possession of 

illegal drugs, or refusal to participate in 

the facility drug and alcohol testing 

program, or a confirmed positive test for 

drugs, drug metabolites, or alcohol in 

violation of the conditions and cutoff 

levels established by § 53.810(f) or the 

consumption of alcoholic beverages 

within the protected area of commercial 

nuclear plants, or a determination of 

unfitness for scheduled work as a result 

of the consumption of alcoholic 

beverages. 

(e) The Commission may require 

information from a generally licensed 

reactor operator to determine whether a 

general license should be revoked or 

suspended with respect to that operator. 

(f) The generally licensed reactor 

operator must not consume or ingest 

alcoholic beverages within the protected 

area of commercial nuclear plants. The 

generally licensed reactor operator must 

not use, possess, or sell any illegal 

drugs. The generally licensed reactor 

operator must not perform activities 

requiring a general license while under 

the influence of alcohol or any 

prescription, over-the-counter, or illegal 

substance that could adversely affect his 

or her ability to safely and competently 

perform these activities. For the purpose 

of this paragraph, with respect to 

alcoholic beverages and drugs, the term 

‘‘under the influence’’ means the 

generally licensed reactor operator 

exceeded, as evidenced by a confirmed 

test result, the lower of the cutoff levels 

for drugs or alcohol contained in 10 CFR 

part 26, or as established by the facility 

licensee. The term ‘‘under the 

influence’’ also means the generally 

licensed reactor operator could be 

mentally or physically impaired as a 

result of substance use including 

prescription and over-the-counter drugs, 

as determined under the provisions, 

policies, and procedures established by 

the facility licensee for its fitness-forduty program, in such a manner as to 

adversely affect his or her ability to 

safely and competently perform 

generally licensed reactor operator 

duties. 

(g) The generally licensed reactor 

operator must notify the Commission 

within 30 days about a conviction for a 

felony. 

§ 53.815 Generally licensed reactor 

operator training, examination, and 

proficiency programs. 

(a) Applicability. The requirements of 

this section apply to each licensee of a 

self-reliant-mitigation facility that has 

not yet certified the permanent 

cessation of operations and permanent 

removal of fuel from the reactor vessel 

as described under § 53.1070. 

(b) Requirements. (1) The licensee 

must develop, implement, and maintain 

training and examination programs that 

demonstrate compliance with the 

requirements of paragraphs (b)(2) and 

(3) of this section. 

(2) The training program must provide 

for both the initial and continuing 

training of generally licensed reactor 

operators and be derived from a systems 

approach to training as defined in this 

part. 

(3)(i) The training program must 

incorporate the instructional 

requirements necessary to provide 

qualified generally licensed reactor 

operators to operate and maintain the 

facility in a safe manner in all modes of 

operation. The training program must 

comply with the facility license, 

including all technical specifications 

and applicable regulations. The facility 

licensee must periodically evaluate and 

revise the training program as 

appropriate to reflect industry 

experience and relevant changes, 

including changes to the facility, 

procedures, regulations, and quality 

assurance (QA) requirements. Facility 

licensee management must periodically 

review the training program for 

effectiveness. 

(ii) The training program must ensure 

that generally licensed reactor operators 

have and maintain the necessary 

knowledge, skills, and abilities. 

(iii) The training program must 

include the generally licensed reactor 

operator manipulating the controls of 

either the facility or a simulation facility 

that demonstrates compliance with the 

requirements of § 53.815(e). 

(iv) The training program must 

include an initial examination program 

for testing a representative sample of the 

knowledge, skills, and abilities needed 

to safely perform generally licensed 

reactor operator duties, to include both 

the examination methods and criteria to 

be used to assess passing performance. 

The facility licensee must provide the 

opportunity for a representative of the 

Commission to be present during initial 

examination administration. 

(v) The training program must include 

a requalification examination program 

for testing a sample of the topics 

included under the systems approach to 

training, to include the examination 

methods and criteria to be used to assess 

passing performance. The 

requalification examination program 

must specify an appropriate periodicity 

for administering a complete 

requalification examination to each 

generally licensed reactor operator, and 

the facility licensee must provide the 

opportunity for a representative of the 

Commission to be present during 

requalification examination 

administration. 

(A) The facility licensee must ensure 

that any generally licensed reactor 

operator who either demonstrates 

unsatisfactory performance on, or fails 

to complete, the requalification 

examination is removed from the 

performance of generally licensed 

reactor operator duties until such time 

that any necessary remedial training has 

been completed and a retake 

examination has been passed. 

(B) [Reserved] 

(vi) The training program must be 

approved by the Commission prior to its 

use, as described under § 53.730(g). The 

examination program must provide for 

valid and reliable examinations and 

must be approved by the Commission 

prior to their use, as described under 

§ 53.730(g). The approved programs are 

subject to the requirements of § 53.1565. 

(c) Records. The following is required 

regarding the documentation of the 

generally licensed reactor operator 

training and examination programs: 

(1) Sufficient records must be 

maintained by the facility licensee to 


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maintain the integrity of the programs 

and kept available for NRC inspection to 

verify the adequacy of the programs. 

(2) The facility licensee must 

maintain records documenting the 

participation of each generally licensed 

reactor operator in the training and 

examination programs. The records 

must contain copies of examinations 

administered, the answers given by the 

generally licensed reactor operator, 

documentation of the grading of 

examinations, and documentation of 

any additional training administered in 

areas in which a generally licensed 

reactor operator exhibited deficiencies. 

The facility licensee must retain these 

records while the associated generally 

licensed reactor operators remain 

employed at the facility. 

(3) Each record required by this part 

must be legible throughout the retention 

period. The record may be the original, 

a reproduced copy, or an electronic 

copy provided that the copy is 

authenticated by authorized personnel. 

(d) Examination integrity. Generally 

licensed reactor operators and facility 

licensees must not engage in any 

activity that compromises the integrity 

of any examination conducted under the 

generally licensed reactor operator 

training and examination programs. The 

integrity of an examination is 

considered compromised if any activity, 

regardless of intent, affected, or, but for 

detection, could have affected the 

equitable and consistent administration 

of the examination. This includes all 

activities related to the preparation, 

administration, and grading of 

examinations. 

(e) Simulation facilities. (1) 

Simulation facilities used for training 

purposes, for maintaining proficiency, 

or for the conduct of examinations must 

demonstrate compliance with the 

following criteria as they relate to the 

facility licensee’s reference plant: 

(i) The simulation facility must be of 

sufficient scope and fidelity for 

individuals to acquire and demonstrate 

the necessary knowledge, skills, and 

abilities to safely perform generally 

licensed reactor operator duties. 

(ii) The simulation facility must 

utilize models relating to nuclear, 

thermal-hydraulic, and other applicable 

design-specific characteristics that 

either replicate the most recent fuel load 

in the reference commercial nuclear 

plant or, prior to initial fuel load (or, for 

a fueled manufactured reactor, prior to 

initiating the physical removal of any 

one of the independent physical 

mechanisms to prevent criticality 

required under § 53.620(d)(1)), replicate 

the intended initial fuel load for the 

reference commercial nuclear plant, 

with the exception of those portions of 

the simulation facility that utilize the 

reference plant itself. 

(iii) Simulator fidelity must be 

demonstrated so that significant control 

manipulations are completed without 

procedural exceptions, simulator 

performance exceptions, or deviation 

from the approved training scenario 

sequence. 

(2) Facility licensees that maintain a 

simulation facility for training purposes, 

for maintaining proficiency, or for the 

conduct of examinations must— 

(i) Conduct performance testing 

throughout the life of the simulation 

facility in a manner sufficient to ensure 

that paragraph (e)(1) of this section is 

met; 

(ii) Retain the results of performance 

testing for 4 years after the completion 

of each performance test or until 

superseded by updated test results; 

(iii) Promptly correct modeling and 

hardware discrepancies and 

discrepancies identified from scenario 

validation and from performance testing 

or provide justification for why the 

presence of such discrepancies will not 

adversely affect the criteria of paragraph 

(e)(1) of this section; 

(iv) Make the results of any 

uncorrected performance test failures 

that may exist at the time of an 

inspection available for NRC review; 

and 

(v) Maintain the provisions for 

examination integrity consistent with 

§ 53.815(d). 

(f) Waiver of examination 

requirement. The facility licensee may 

waive any or all of the requirements for 

an examination in accordance with the 

facility licensee’s Commission-approved 

generally licensed reactor operator 

training and examination programs. 

(g) Proficiency. The facility licensee 

must develop, implement, and maintain 

a proficiency program to allow generally 

licensed reactor operators to maintain 

proficiency regarding position functions 

and familiarity with plant status. This 

program must include those steps that 

will be taken in order to re-establish 

proficiency when it cannot be 

maintained. 

§ 53.820 Cessation of individual 

applicability. 

The general license ceases to be 

applicable on an individual basis once 

a generally licensed reactor operator is 

no longer being employed in a position 

that may involve the manipulation of 

the controls of the self-reliant mitigation 

facility. 

§ 53.830 Training and qualification of 

commercial nuclear plant personnel. 

(a) This section addresses personnel 

training requirements. The regulations 

within this section are applicable to all 

applicants for or holders of OLs or COLs 

under this part. 

(b) Prior to initial fuel load (or, for a 

fueled manufactured reactor, prior to 

initiating the physical removal of any 

one of the independent physical 

mechanisms to prevent criticality 

required under § 53.620(d)(1)), each 

holder of an operating or COL under 

this part must, with sufficient time to 

provide trained and qualified personnel 

to operate the facility, establish, 

implement, and maintain a training 

program that demonstrates compliance 

with the requirements of paragraphs (c) 

and (d) of this section. 

(c) The training program must be 

derived from a systems approach to 

training as defined in this part and must 

provide, at a minimum, for the training 

and qualification of the following 

categories of commercial nuclear plant 

personnel: 

(1) Supervisors (e.g., shift 

supervisors); 

(2) Technicians (e.g., maintenance, 

chemistry, and radiological); and 

(3) Other appropriate operating 

personnel (e.g., auxiliary operators, 

certified fuel handlers, and individuals 

who provide engineering expertise to 

on-shift operating personnel). 

(d) The training program must 

incorporate the instructional 

requirements necessary to provide 

qualified personnel to operate 

components of a commercial nuclear 

plant and maintain the facility in a safe 

manner in all modes of operation. The 

training program must be developed to 

be in compliance with the facility 

license, including all technical 

specifications and applicable 

regulations. 

(1) The training program must be 

periodically evaluated and revised as 

appropriate to reflect industry 

experience and relevant changes, 

including changes to the facility, 

procedures, regulations, and QA 

requirements. The training program 

must be periodically reviewed by 

facility licensee management for 

effectiveness. 

(2) Sufficient records must be 

maintained by the facility licensee to 

maintain program integrity and kept 

available for NRC inspection to verify 

the adequacy of the training program. 

§ 53.845 Programs. 

(a) The required plant programs under 

this part must include but are not 

necessarily limited to the programs 


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described in the following sections of 

this subpart. Licensees may combine, 

separate, and otherwise organize 

programs and related documents as 

appropriate for the technologies and 

organizations associated with the 

commercial nuclear plant. 

(b) In addition to the programs 

described in the following sections, 

programs must be provided for each 

commercial nuclear plant, if necessary, 

to ensure that the performance of design 

features and human actions are 

consistent with the analyses performed 

under §§ 53.450 and 53.730 and that the 

plant will demonstrate compliance with 

the safety criteria defined in §§ 53.210 

and 53.220. 

§ 53.850 Radiation protection. 

(a) Each holder of an OL or COL 

under this part must develop, 

implement, and maintain a Radiation 

Protection Program for operations that is 

commensurate with the scope and 

extent of licensed activities under this 

part and includes measures for limiting 

and monitoring radioactive plant 

effluents and limiting and monitoring 

the dose to individuals working with 

radioactive materials in accordance with 

10 CFR part 20. 

(b) Each holder of an OL or COL 

under this part must develop, 

implement, and maintain a program for 

the control of radioactive effluents and 

for keeping the doses to members of the 

public from radioactive effluents as low 

as is reasonably achievable and for 

environmental monitoring. The program 

must be contained in an Offsite Dose 

Calculations Manual, must be 

implemented by procedures, and must 

include remedial actions to be taken 

whenever the program limits are 

exceeded. The Offsite Dose Calculations 

Manual must— 

(1) Contain the methodology and 

parameters used in the calculation of 

offsite doses resulting from radioactive 

gaseous and liquid effluents, in the 

calculation of gaseous and liquid 

effluent monitoring alarm and trip 

setpoints, and in the conduct of the 

radiological environmental monitoring 

program; and 

(2) Contain the radioactive effluent 

controls and radiological environmental 

monitoring activities, and descriptions 

of the information that should be 

included in the Annual Radiological 

Environmental Operating and 

Radioactive Effluent Release Reports 

required by § 53.1645. 

(c) Each holder of an OL or COL 

under this part must develop, 

implement, and maintain a Process 

Control Program that identifies the 

administrative and operational controls 

for solid radioactive waste processing, 

process parameters, and surveillance 

requirements sufficient to ensure 

compliance with the requirements of 10 

CFR part 20, 10 CFR part 61, and 10 

CFR part 71. 

§ 53.855 Emergency preparedness. 

(a) Each holder of an OL or COL 

under this part must have an emergency 

response plan that must contain 

information needed to demonstrate 

compliance with either the 

requirements in § 50.160 of this chapter 

or the requirements in appendix E to 

part 50 and the planning standards of 

§ 50.47(b) of this chapter. 

(b) No initial OL, initial COL, or early 

site permit that includes complete and 

integrated emergency plans will be 

issued under this part unless a finding 

is made by the NRC, in accordance with 

§ 50.47 of this chapter, that there is 

reasonable assurance that adequate 

protective measures can and will be 

taken in the event of a radiological 

emergency. 

§ 53.860 Security programs. 

(a) Physical Protection Program. Each 

holder of an OL or COL under this part 

must develop, implement, and maintain 

a physical protection program under the 

following requirements: 

(1) The licensee must implement 

security requirements for the protection 

of special nuclear material based on the 

type, enrichment, and quantity in 

accordance with 10 CFR part 73, as 

applicable, and implement security 

requirements for the protection of 

Category 1 and Category 2 quantities of 

radioactive material in accordance with 

10 CFR part 37, as applicable; and 

(2) The licensee must demonstrate 

compliance with the provisions set forth 

in either §§ 73.55 or 73.100 of this 

chapter, unless the licensee 

demonstrates compliance with the 

following criterion: 

(i) The radiological consequences 

from a design-basis threat-initiated 

event involving the loss of engineered 

systems for decay heat removal and 

possible breaches in physical structures 

surrounding the reactor, spent fuel, and 

other inventories of radioactive 

materials result in offsite doses below 

the values in § 53.210. 

(ii) The applicant must perform a sitespecific analysis, including 

identification of target sets, to 

demonstrate that the criterion in 

§ 53.860(a)(2)(i) is satisfied. The analysis 

must assume that licensee mitigation 

and recovery actions, including any 

operator actions, are unavailable or 

ineffective. The licensee must maintain 

the analysis until the permanent 

cessation of operations and permanent 

removal of fuel from the reactor vessel 

as described under § 53.1070. 

(b) Fitness for Duty. Each holder of an 

OL or COL under this part must 

develop, implement, and maintain a 

fitness for duty program under 10 CFR 

part 26. 

(c) Access Authorization. Each holder 

of an OL or COL under this part must 

develop, implement, and maintain an 

access authorization program under 

§ 73.120 of this chapter if the criterion 

in § 53.860(a)(2)(i) is satisfied, or the 

requirements in § 73.56 of this chapter 

if the criterion is not satisfied. 

(d) Cybersecurity. Each holder of an 

OL or COL under this part must 

develop, implement, and maintain a 

cybersecurity program under §§ 73.54 or 

73.110 of this chapter. 

(e) Information Security. Each holder 

of an OL or COL under this part must 

develop, implement, and maintain an 

information protection system under 

§§ 73.21, 73.22, and 73.23 of this 

chapter, as applicable. 

§ 53.865 Quality assurance. 

Each holder of an OL or COL under 

this part must develop, implement, and 

maintain a quality assurance program in 

accordance with appendix B of part 50 

of this chapter. A written quality 

assurance program manual must be 

developed and used to guide the 

conduct of the program in accordance 

with generally accepted consensus 

codes and standards that have been 

endorsed or otherwise found acceptable 

by the NRC. 

§ 53.870 Integrity assessment programs. 

Each holder of an OL or COL under 

this part must develop, implement, and 

maintain an integrity assessment 

program to monitor, evaluate, and 

manage— 

(a) The effects of plant aging on SR 

and NSRSS SSCs. The program may 

refer to surveillances, tests, and 

inspections conducted for specific SSCs 

in accordance with other requirements 

in this part or conducted in accordance 

with applicable consensus codes and 

standards endorsed or otherwise found 

acceptable by the NRC; 

(b) Cyclic or transient load limits to 

ensure that SR and NSRSS SSCs are 

maintained within the applicable design 

limits; and 

(c) Degradation mechanisms related to 

chemical interactions, operating 

temperatures, effects of irradiation, and 

other environmental factors to ensure 

that the capabilities, availability, and 

reliability of SR and NSRSS SSCs 

demonstrate compliance with the 


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functional design criteria of §§ 53.410 

and 53.420. 

§ 53.875 Fire protection. 

(a)(1) Each holder of an OL or COL 

under this part must have a fire 

protection plan that describes the 

overall fire protection program for the 

facility; identifies the various positions 

within the licensee’s organization that 

are responsible for the program; states 

the authorities that are delegated to each 

of these positions to implement those 

responsibilities; and outlines the plans 

for fire protection, fire detection and 

suppression capability; and limitation of 

fire damage. 

(2) The fire protection plan must also 

describe specific features necessary to 

implement the program described in 

paragraph (a)(1) of this section such as 

the following: administrative controls 

and personnel requirements for fire 

prevention and manual fire suppression 

activities; automatic and manually 

operated fire detection and suppression 

systems; and the means to limit fire 

damage to SSCs so that the capability to 

demonstrate compliance with the 

requirements of § 53.210 is ensured. 

(b)(1) Each holder of an OL or COL 

under this part must develop a 

performance-based or deterministic fire 

protection program that demonstrates 

compliance with the safety criteria 

outlined in §§ 53.210 and 53.220, 

related safety functions outlined in 

§ 53.230, and defense in depth as 

outlined in § 53.250 with specific fire 

protection measures related to fire 

prevention, fire detection, and fire 

suppression. 

(2) The fire protection program must 

comply with the following: 

(i) Safety-related and NSRSS SSCs 

must be designed, located, and 

maintained to minimize, consistent with 

other safety requirements, the 

probability and effect of fires and 

explosions. 

(ii) Noncombustible and fire-resistant 

materials must be used wherever 

practical throughout the facility, 

particularly in locations with SR and 

NSRSS SSCs. 

(iii) Fire detection and fire 

suppression systems of appropriate 

capacity and capability must be 

provided and designed and maintained 

to minimize the adverse effects of fires 

on SR and NSRSS SSCs. 

(iv) Fire suppression systems must be 

designed and maintained to ensure that 

their rupture or inadvertent operation 

does not significantly impair the ability 

of SR and NSRSS SSCs to perform their 

safety functions to satisfy § 53.230. 

§ 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. 

§ 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 

§ 53.715; 

(3) Program requirements under this 

subpart F of this part; 

(4) Emergency operating procedures, 

if developed to address the role of 

human actions in responding to LBEs; 

(5) Accident management guidelines, 

if developed to address the role of 

human actions in responding to LBEs; 

(6) Procedures for each area in which 

licensed special nuclear material is 

handled, used, or stored to protect 

personnel upon the sounding of a 

criticality alarm required by 

§ 53.440(m); and 

(7) Procedures that describe how the 

licensee will address the following areas 

if the licensee is notified of a potential 

aircraft threat: 

(i) Verification of the authenticity of 

threat notifications; 

(ii) Maintenance of continuous 

communication with threat notification 

sources; 

(iii) Contacting all onsite personnel 

and applicable offsite response 

organizations; 

(iv) Onsite actions necessary to 

enhance the capability of the facility to 

mitigate the consequences of an aircraft 

impact; 

(v) Measures to reduce visual 

discrimination of the site relative to its 

surroundings or individual buildings 

within the protected area; 

(vi) Dispersal of equipment and 

personnel, as well as rapid entry into 

site protected areas for essential onsite 

personnel and offsite responders who 

are necessary to mitigate the event; and 

(vii) Recall of site personnel. 

Subpart G—Decommissioning 

Requirements 

§ 53.1000 Scope and purpose. 

This subpart defines the requirements 

related to decommissioning for 

applicants for, or holders of, an 

operating license (OL) or combined 

license (COL). The requirements related 

to maintaining financial assurance for 

decommissioning are in §§ 53.1010 

through 53.1060. The requirements for 

transitioning from operations to 

decommissioning and for the release of 

property and termination of the license 

are in §§ 53.1070 through 53.1080. 


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§ 53.1010 Financial assurance for 

decommissioning. 

(a) This section establishes 

requirements for indicating to the U.S. 

Nuclear Regulatory Commission (NRC) 

how an applicant for or holder of an OL 

or COL under this part will provide 

reasonable assurance that funds will be 

available for the decommissioning 

process. Reasonable assurance consists 

of a series of steps as provided in 

paragraph (b) of this section and 

§§ 53.1020, 53.1030 and 53.1040. 

Funding for the decommissioning of 

commercial nuclear plants may also be 

subject to the regulation of Federal or 

State government agencies (e.g., Federal 

Energy Regulatory Commission (FERC) 

and State Public Utility Commissions) 

that have jurisdiction over rate 

regulation. The requirements of this 

subpart, in particular § 53.1020, are in 

addition to, and not a substitution for, 

other requirements, and are not 

intended to be used by themselves or by 

other agencies to establish rates. 

(b) Each applicant for an OL or COL 

under this part must prepare a plan and 

an associated decommissioning report 

that ensures and documents that 

adequate funding will be available to 

decommission the facility. Each holder 

of an OL or COL must implement and 

maintain the plan. 

(1)(i) Before the Commission issues an 

OL under this part, the applicant must 

update the decommissioning report to 

certify that it has provided financial 

assurance for decommissioning in the 

amount proposed in the application and 

approved by the NRC under § 53.1020. 

(ii) No later than 30 days after the 

Commission issues the notice of 

intended operation under § 53.1452 for 

a COL under this part, the licensee must 

update the decommissioning report to 

certify that it has provided financial 

assurance for decommissioning in the 

amount proposed in the application and 

approved by the NRC under § 53.1020. 

(2) The amount of financial assurance 

for decommissioning to be provided 

must be based on a site-specific cost 

estimate for decommissioning the 

facility under § 53.1020. 

(3) The amount of financial assurance 

for decommissioning to be provided 

must be adjusted annually using a rate 

at least equal to that stated in § 53.1030. 

(4) The amount of financial assurance 

for decommissioning to be provided 

must be covered by one or more of the 

methods described in § 53.1040 as 

acceptable to the NRC. A copy of the 

financial instrument obtained to satisfy 

the requirements of § 53.1040 must be 

submitted to the NRC as part of the 

application for an OL under this part; 

however, an applicant for or holder of 

a COL need not obtain such financial 

instrument or submit a copy to the 

Commission except as provided in 

§ 53.1060(b). 

§ 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 § 53.1030. Finally, sitespecific 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. 

§ 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 

§ 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.’’ 

§ 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 § 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 

§ 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 


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estimate under § 53.1020 may take 

credit for projected earnings on the 

external sinking funds using up to a 2 

percent annual real rate of return from 

the time of future funds’ collection 

through the time of termination of the 

license. A licensee may use a credit of 

greater than 2 percent if the licensee’s 

rate-setting authority has specifically 

authorized a higher rate. Actual 

earnings on existing funds may be used 

to calculate future fund needs. A 

licensee whose rates for 

decommissioning costs cover only a 

portion of these costs may make use of 

this method only for the portion of these 

costs that are collected in one of the 

manners described in this paragraph. 

This method may be used as the 

exclusive mechanism relied upon for 

providing financial assurance for 

decommissioning in the following 

circumstances: 

(1) By a licensee that recovers, either 

directly or indirectly, the estimated total 

cost of decommissioning through rates 

established by ‘‘cost of service’’ or 

similar ratemaking regulation. Public 

utility districts, municipalities, rural 

electric cooperatives, and State and 

Federal agencies, including associations 

of any of the foregoing, that establish 

their own rates and are able to recover 

their cost of service allocable to 

decommissioning, are deemed to satisfy 

this condition. 

(2) By a licensee whose source of 

revenues for its external sinking fund is 

a ‘‘non-bypassable charge,’’ the total 

amount of which will provide funds 

estimated to be needed for 

decommissioning pursuant to 

§§ 53.1020, 53.1060, or 53.1575. 

(c) A surety method, insurance, or 

other guarantee method. (1) These 

methods guarantee that 

decommissioning costs will be paid. A 

surety method may be in the form of a 

surety bond, or letter of credit. Any 

surety method or insurance used to 

provide financial assurance for 

decommissioning must contain the 

following conditions: 

(i) The surety method or insurance 

must be open-ended, or, if written for a 

specified term, such as 5 years, must be 

renewed automatically, unless 90 days 

or more prior to the renewal day the 

issuer notifies the NRC, the beneficiary, 

and the licensee of its intention not to 

renew. The surety or insurance must 

also provide that the full-face amount be 

paid to the beneficiary automatically 

prior to the expiration without proof of 

forfeiture if the licensee fails to provide 

a replacement acceptable to the NRC 

within 30 days after receipt of 

notification of cancellation. 

(ii) The surety or insurance must be 

payable to a trust established for 

decommissioning costs. The trustee and 

trust must be acceptable to the NRC. An 

acceptable trustee includes an 

appropriate State or Federal 

Government agency or an entity that has 

the authority to act as a trustee and 

whose trust operations are regulated and 

examined by a Federal or State agency. 

(2) A parent company guarantee of 

funds for decommissioning costs based 

on a financial test may be used if the 

guarantee and test are as contained in 

appendix A to 10 CFR part 30. 

(3) For commercial companies that 

issue bonds, a guarantee of funds by the 

applicant or licensee for 

decommissioning costs based on a 

financial test may be used if the 

guarantee and test are as contained in 

appendix C to 10 CFR part 30. For 

commercial companies that do not issue 

bonds, a guarantee of funds by the 

applicant or licensee for 

decommissioning costs may be used if 

the guarantee and test are as contained 

in appendix D to 10 CFR part 30. A 

guarantee by the applicant or licensee 

may not be used in any situation in 

which the applicant or licensee has a 

parent company holding majority 

control of voting stock of the company. 

(d) Funding method for Federal 

licensees. For a Federal licensee, a 

statement of intent containing a cost 

estimate for decommissioning and 

indicating that funds for 

decommissioning will be obtained when 

necessary. 

(e) Contractual funding method. 

Contractual obligation(s) on the part of 

a licensee’s customer(s), the total 

amount of which over the duration of 

the contract(s) will provide the 

licensee’s total share of uncollected 

funds estimated to be needed for 

decommissioning pursuant to 

§§ 53.1020, 53.1060, or 53.1575. To be 

acceptable to the NRC as a method of 

decommissioning funding assurance, 

the terms of the contract(s) must include 

provisions that the buyer(s) of electricity 

or other products will pay for the 

decommissioning obligations specified 

in the contract(s), notwithstanding the 

operational status either of the licensed 

plant to which the contract(s) pertains 

or force majeure provisions. All 

proceeds from the contract(s) for 

decommissioning funding will be 

deposited to the external sinking fund. 

The NRC reserves the right to evaluate 

the terms of any contract(s) and the 

financial qualifications of the 

contracting entity or entities offered as 

assurance for decommissioning funding. 

(f) Other funding mechanisms. Any 

other mechanism, or combination of 

mechanisms, that provides, as 

determined by the NRC upon its 

evaluation of the specific circumstances 

of each licensee submittal, assurance of 

decommissioning funding equivalent to 

that provided by the mechanisms 

specified in paragraphs (a) through (e) of 

this section. Licensees who do not have 

sources of funding described in 

paragraph (b) of this section may use an 

external sinking fund in combination 

with a guarantee mechanism, as 

specified in paragraph (c) of this 

section, provided that the total amount 

of funds estimated to be necessary for 

decommissioning is assured. 

§ 53.1045 Limitations on the use of 

decommissioning trust funds. 

(a)(1) Decommissioning trust funds 

may be used by licensees if— 

(i) The withdrawals are for expenses 

for decommissioning activities 

consistent with the definition of 

decommission or decommissioning in 

§ 53.020; 

(ii) The expenditure would not reduce 

the value of the decommissioning trust 

below an amount necessary to place and 

maintain the reactor in a safe storage 

condition if unforeseen conditions or 

expenses arise; and 

(iii) The withdrawals would not 

inhibit the ability of the licensee to 

complete funding of any shortfalls in 

the decommissioning trust needed to 

ensure the availability of funds to 

ultimately release the site and terminate 

the license. 

(2) Initially, 3 percent of the amount 

determined in accordance with 

§ 53.1020 may be used for 

decommissioning planning. For 

licensees that have submitted the 

certifications required under § 53.1070 

and commencing 90 days after the NRC 

has received the post-shutdown 

decommissioning activities report 

(PSDAR) required by § 53.1060, an 

additional 20 percent may be used. An 

updated site-specific DCE must be 

submitted to the NRC prior to the 

licensee using any funding in excess of 

these amounts. 

(b) Licensees that are not ‘‘electric 

utilities’’ as defined in § 53.020 that use 

prepayment or an external sinking fund 

to provide financial assurance must 

provide in the terms of the arrangements 

governing the trust, escrow account, or 

Government fund, used to segregate and 

manage the funds that— 

(1) The trustee, manager, investment 

advisor, or other person directing 

investment of the funds— 

(i) Is prohibited from investing the 

funds in securities or other obligations 

of the licensee or any other owner or 

operator of any commercial nuclear 


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plant or their affiliates, subsidiaries, 

successors or assigns, or in a mutual 

fund in which at least 50 percent of the 

fund is invested in the securities of a 

licensee or parent company whose 

subsidiary is an owner or operator of a 

foreign or domestic commercial nuclear 

plant. However, the funds may be 

invested in securities tied to market 

indices or other non-nuclear sector 

collective, commingled, or mutual 

funds, provided that no more than 10 

percent of trust assets may be indirectly 

invested in securities of any entity 

owning or operating one or more 

commercial nuclear plants. 

(ii) Is obligated at all times to adhere 

to a standard of care set forth in the 

trust, which either shall be the standard 

of care, whether in investing or 

otherwise, required by State or Federal 

law or one or more State or Federal 

regulatory agencies with jurisdiction 

over the trust funds, or, in the absence 

of any such standard of care, whether in 

investing or otherwise, that a prudent 

investor would use in the same 

circumstances. The term ‘‘prudent 

investor,’’ shall have the same meaning 

as set forth in FERC’s ‘‘Regulations 

Governing Nuclear Plant 

Decommissioning Trust Funds’’ at 18 

CFR 35.32(a)(3), or any successor 

regulation. 

(2) The licensee, its affiliates, and its 

subsidiaries are prohibited from being 

engaged as investment manager for the 

funds or from giving day-to-day 

management direction of the funds’ 

investments or direction on individual 

investments by the funds, except in the 

case of passive fund management of 

trust funds where management is 

limited to investments tracking market 

indices. 

(3) The trust, escrow account, 

Government fund, or other account used 

to segregate and manage the funds may 

not be amended in any material respect 

without written notification to the 

Director, Office of Nuclear Reactor 

Regulation, or Director, Office of 

Nuclear Material Safety and Safeguards, 

as applicable, at least 30 working days 

before the proposed effective date of the 

amendment. The licensee must provide 

the text of the proposed amendment and 

a statement of the reason for the 

proposed amendment. The trust, escrow 

account, Government fund, or other 

account may not be amended if the 

person responsible for managing the 

trust, escrow account, Government 

fund, or other account receives written 

notice of objection from the Director, 

Office of Nuclear Reactor Regulation, or 

Director, Office of Nuclear Material 

Safety and Safeguards, as applicable, 

within the notice period. 

(4) Except for withdrawals being 

made under paragraph (a) of this section 

or for payments of ordinary 

administrative costs (including taxes) 

and other incidental expenses of the 

fund (including legal, accounting, 

actuarial, and trustee expenses) in 

connection with the operation of the 

fund, no disbursement or payment may 

be made from the trust, escrow account, 

Government fund, or other account used 

to segregate and manage the funds until 

written notice of the intention to make 

a disbursement or payment has been 

given to the Director, Office of Nuclear 

Reactor Regulation, or Director, Office of 

Nuclear Material Safety and Safeguards, 

as applicable, at least 30 working days 

before the date of the intended 

disbursement or payment. The 

disbursement or payment from the trust, 

escrow account, Government fund or 

other account may be made following 

the 30 working day notice period if the 

person responsible for managing the 

trust, escrow account, Government 

fund, or other account does not receive 

written notice of objection from the 

Director, Office of Nuclear Reactor 

Regulation, or Director, Office of 

Nuclear Material Safety and Safeguards, 

as applicable, within the notice period. 

Disbursements or payments from the 

trust, escrow account, Government 

fund, or other account used to segregate 

and manage the funds, other than for 

payment of ordinary administrative 

costs (including taxes) and other 

incidental expenses of the fund 

(including legal, accounting, actuarial, 

and trustee expenses) in connection 

with the operation of the fund, are 

restricted to decommissioning expenses 

or transfer to another financial 

assurance method acceptable under 

§ 53.1040 until final decommissioning 

has been completed. After 

decommissioning has begun and 

withdrawals from the decommissioning 

fund are made under paragraph (a) of 

this section, no further notification need 

be made to the NRC. 

(c) Licensees that are ‘‘electric 

utilities’’ under § 53.020 that use 

prepayment or an external sinking fund 

to provide financial assurance must 

include a provision in the terms of the 

trust, escrow account, Government 

fund, or other account used to segregate 

and manage funds that except for 

withdrawals being made under 

paragraph (a) of this section or for 

payments of ordinary administrative 

costs (including taxes) and other 

incidental expenses of the fund 

(including legal, accounting, actuarial, 

and trustee expenses) in connection 

with the operation of the fund, no 

disbursement or payment may be made 

from the trust, escrow account, 

Government fund, or other account used 

to segregate and manage the funds until 

written notice of the intention to make 

a disbursement or payment has been 

given the Director, Office of Nuclear 

Reactor Regulation, or Director, Office of 

Nuclear Material Safety and Safeguards, 

as applicable, at least 30 working days 

before the date of the intended 

disbursement or payment. The 

disbursement or payment from the trust, 

escrow account, Government fund or 

other account may be made following 

the 30 working day notice period if the 

person responsible for managing the 

trust, escrow account, Government 

fund, or other account does not receive 

written notice of objection from the 

Director, Office of Nuclear Reactor 

Regulation, or Director, Office of 

Nuclear Material Safety and Safeguards, 

as applicable, within the notice period. 

Disbursements or payments from the 

trust, escrow account, Government 

fund, or other account used to segregate 

and manage the funds, other than for 

payment of ordinary administrative 

costs (including taxes) and other 

incidental expenses of the fund 

(including legal, accounting, actuarial, 

and trustee expenses) in connection 

with the operation of the fund, are 

restricted to decommissioning expenses 

or transfer to another financial 

assurance method acceptable under 

§ 53.1040 until final decommissioning 

has been completed. After 

decommissioning has begun and 

withdrawals from the decommissioning 

fund are made under paragraph (a) of 

this section, no further notification need 

be made to the NRC. 

(d) A licensee that is not an ‘‘electric 

utility’’ under § 53.020 and using a 

surety method, insurance, or other 

guarantee method to provide financial 

assurance must provide that the trust 

established for decommissioning costs 

to which the surety or insurance is 

payable contains in its terms the 

requirements in § 53.1045(b)(1) through 

(4). 

§ 53.1050 NRC oversight. 

The NRC reserves the right to take the 

following steps in order to ensure a 

licensee’s adequate accumulation of 

decommissioning funds: review, as 

needed, the rate of accumulation of 

decommissioning funds and, either 

independently or in cooperation with 

FERC and the licensee’s State Public 

Utility Commission, take additional 

actions as appropriate on a case-by-case 

basis, including modification of a 

licensee’s schedule for the accumulation 

of decommissioning funds. 


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§ 53.1060 Reporting and recordkeeping 

requirements. 

(a) Each holder of an OL under this 

part or holder of a COL under this part 

after the date that the Commission has 

made the finding under § 53.1452(g) 

must report, at least once every 2 years, 

by March 31, on the status of its 

certification of decommissioning 

funding for each commercial nuclear 

reactor or part of a commercial nuclear 

reactor that it owns. The information in 

this report must include, at a minimum, 

the amount of decommissioning funds 

estimated to be required under 

§§ 53.1020 and 53.1030; the amount of 

decommissioning funds accumulated to 

the end of the calendar year preceding 

the date of the report; a schedule of the 

annual amounts remaining to be 

collected; the assumptions used 

regarding rates of escalation in 

decommissioning costs, rates of 

earnings on decommissioning funds, 

and rates of other factors used in 

funding projections; any contracts upon 

which the licensee is relying under 

§ 53.1040(e); any modifications 

occurring to a licensee’s method of 

providing financial assurance since the 

last submitted report; and any material 

changes to trust agreements. If any of 

the preceding items is not applicable, 

the licensee should so state in its report. 

Any licensee for a plant that is within 

5 years of the projected end of its 

operation, or where conditions have 

changed such that it will close within 5 

years (before the end of its licensed life), 

or that has already closed (before the 

end of its licensed life), or that is 

involved in a merger or an acquisition 

must submit this report annually. 

(b) Each holder of a COL under this 

part must, 2 years before and 1 year 

before the scheduled date for initial 

loading of fuel (or, for a fueled 

manufactured reactor, 2 years before 

and 1 year before the scheduled date for 

initiating the physical removal of any 

one of the independent physical 

mechanisms to prevent criticality 

required under § 53.620(d)(1)), submit a 

report to the NRC containing a 

certification updating the DCEs and a 

copy of the financial instrument to be 

used to satisfy § 53.1040. No later than 

30 days after the Commission publishes 

notice in the Federal Register under 

§ 53.1452(a), the licensee must submit 

an updated decommissioning report 

required under § 53.1010(b)(1)(ii), 

including a copy of the financial 

instrument obtained to satisfy § 53.1040. 

(c) Each licensee must keep records of 

information important to the safe and 

effective decommissioning of the facility 

in an identified location until the 

license is terminated by the 

Commission. If records of relevant 

information are kept for other purposes, 

reference to these records and their 

locations may be used. Information the 

Commission considers important to 

decommissioning consists of— 

(1) Records of spills or other unusual 

occurrences involving the spread of 

contamination in and around the 

facility, equipment, or site. These 

records may be limited to instances 

when significant contamination remains 

after any cleanup procedures or when 

there is reasonable likelihood that 

contaminants may have spread to 

inaccessible areas as in the case of 

possible seepage into porous materials 

such as concrete. These records must 

include any known information on 

identification of involved nuclides, 

quantities, forms, and concentrations. 

(2) As-built drawings and 

modifications of structures and 

equipment in restricted areas where 

radioactive materials are used and/or 

stored and of locations of possible 

inaccessible contamination such as 

buried pipes that may be subject to 

contamination. If required drawings are 

referenced, each relevant document 

need not be indexed individually. If 

drawings are not available, the licensee 

must substitute appropriate records of 

available information concerning these 

areas and locations. 

(3) Records of the cost estimate 

performed for the decommissioning 

funding plan or of the amount certified 

for decommissioning, and records of the 

funding method used for assuring funds 

if either a funding plan or certification 

is used. 

(4) Records of— 

(i) The licensed site area, as originally 

licensed and any revisions, which must 

include a site map and any acquisition 

or use of property outside the originally 

licensed site area for the purpose of 

receiving, possessing, or using licensed 

materials; 

(ii) The licensed activities carried out 

on the acquired or used property; and 

(iii) The release and final disposition 

of any property recorded in paragraph 

(c)(4)(i) of this section, the historical site 

assessment performed for the release, 

radiation surveys performed to support 

release of the property, submittals to the 

NRC made under § 53.1070, and the 

methods employed to ensure that the 

property met the radiological criteria of 

subpart E of 10 CFR part 20 at the time 

the property was released. 

(d) Each holder of an OL or COL 

under this part must at or about 5 years 

prior to the projected end of operations 

submit a preliminary DCE which 

includes an up-to-date assessment of the 

major factors that could affect the cost 

to decommission. 

(e) Prior to or within 2 years following 

permanent cessation of operations, the 

licensee must submit a PSDAR to the 

NRC, and a copy to the affected State(s). 

The PSDAR must contain a description 

of the planned decommissioning 

activities along with a schedule for their 

accomplishment, a discussion that 

provides the reasons for concluding that 

the environmental impacts associated 

with site-specific decommissioning 

activities will be bounded by 

appropriate previously issued 

environmental impact statements, and a 

site-specific DCE, including the 

projected cost of managing irradiated 

fuel. 

(f) For decommissioning activities 

that delay completion of 

decommissioning by including a period 

of storage or surveillance, the licensee 

must provide a means of adjusting cost 

estimates and associated funding levels 

over the storage or surveillance period. 

(g) After submitting its site-specific 

DCE required by paragraph (e) of this 

section, and until the licensee has 

completed its final radiation survey and 

demonstrated that residual radioactivity 

has been reduced to a level that permits 

termination of its license, the licensee 

must annually submit to the NRC, by 

March 31, a financial assurance status 

report. The report must include the 

following information, current through 

the end of the previous calendar year: 

(1) The amount spent on 

decommissioning, both cumulative and 

over the previous calendar year, the 

remaining balance of any 

decommissioning funds, and the 

amount provided by other financial 

assurance methods being relied upon; 

(2) An estimate of the costs to 

complete decommissioning, reflecting 

any difference between actual and 

estimated costs for work performed 

during the year, and the 

decommissioning criteria upon which 

the estimate is based; 

(3) Any modifications occurring to a 

licensee’s current method of providing 

financial assurance since the last 

submitted report; and 

(4) Any material changes to trust 

agreements or financial assurance 

contracts. 

(5) If the sum of the balance of any 

remaining decommissioning funds, plus 

earnings on such funds calculated at not 

greater than a 2 percent real rate of 

return, together with the amount 

provided by other financial assurance 

methods being relied upon, does not 

cover the estimated cost to complete the 

decommissioning, the financial 

assurance status report must include 


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additional financial assurance to cover 

the estimated cost of completion. 

(h) After submitting its site-specific 

DCE required by paragraph (e) of this 

section, the licensee must annually 

submit to the NRC, by March 31, a 

report on the status of its funding for 

managing irradiated fuel. The report 

must include the following information, 

current through the end of the previous 

calendar year: 

(1) The amount of funds accumulated 

to cover the cost of managing the 

irradiated fuel; 

(2) The projected cost of managing 

irradiated fuel until title to the fuel and 

possession of the fuel is transferred to 

the Secretary of Energy; and 

(3) If the funds accumulated do not 

cover the projected cost, a plan to obtain 

additional funds to cover the cost. 

§ 53.1070 Termination of license. 

For each holder of an OL or COL 

under this part— 

(a)(1) When the licensee has 

determined to permanently cease 

operations the licensee must, within 30 

days, submit a written certification to 

the NRC, consistent with the 

requirements of § 53.040(b)(8); 

(2) When appropriate to support 

decommissioning activities and the 

eventual permanent removal of fuel 

from the reactor vessel, the licensee 

must develop defueled technical 

specifications by reviewing the 

operational technical specifications and 

determining which specifications no 

longer apply during decommissioning 

and which ones should remain 

applicable. The licensee must make the 

appropriate submittals to the NRC in 

accordance with § 53.1510 to request 

changes to the technical specifications; 

and 

(3)(i) Once fuel has been permanently 

removed from the reactor vessel, the 

licensee must submit a written 

certification to the NRC that meets the 

requirements of § 53.040(b)(9); and 

(ii) The licensee must establish and 

maintain staffing consisting of certified 

fuel handlers, as defined under § 53.020, 

and other non-licensed personnel with 

appropriate qualifications, and in 

sufficient numbers, to ensure support 

for facility operations and radiological 

control activities, as required by the 

facility defueled technical 

specifications. These personnel must be 

subject to the training requirements of 

§ 53.830. 

(b) Upon docketing of the 

certifications for permanent cessation of 

operations and permanent removal of 

fuel from the reactor vessel, or when a 

final legally effective order to 

permanently cease operations has come 

into effect, the license issued under this 

part no longer authorizes operation of 

the reactor or emplacement or retention 

of fuel into the reactor vessel. 

(c) Decommissioning will be 

completed within 60 years of permanent 

cessation of operations. Completion of 

decommissioning beyond 60 years will 

be approved by the Commission only 

when necessary to protect public health 

and safety. Factors that will be 

considered by the Commission in 

evaluating an alternative that provides 

for completion of decommissioning 

beyond 60 years of permanent cessation 

of operations include unavailability of 

waste disposal capacity and other sitespecific factors affecting the licensee’s 

capability to carry out 

decommissioning, including presence of 

other nuclear facilities at the site. 

(d)(1) Prior to or within 2 years 

following permanent cessation of 

operations, the licensee must submit a 

PSDAR and site-specific DCE in 

accordance with § 53.1060(e). 

(2) The NRC must notice receipt of the 

PSDAR and make the PSDAR publicly 

available and publish notice of its 

availability for public comment in the 

Federal Register. The NRC must also 

schedule a public meeting readily 

accessible to individuals in the vicinity 

of the licensee’s facility. The NRC must 

publish a notice in the Federal Register 

and in a forum, such as local 

newspapers, that is readily accessible to 

individuals in the vicinity of the site, 

announcing the date, time, and location 

of the meeting, along with a brief 

description of the purpose of the 

meeting. 

(e) Licensees must not perform any 

major decommissioning activities, as 

defined in § 53.020, until 90 days after 

the NRC has received the licensee’s 

PSDAR submittal and until 

certifications of permanent cessation of 

operations and permanent removal of 

fuel from the reactor vessel, as required 

under paragraph (a) of this section, have 

been submitted. 

(f) Licensees must not perform any 

decommissioning activities, as defined 

in § 53.020, that— 

(1) Foreclose release of the site for 

possible unrestricted use; 

(2) Result in significant 

environmental impacts not previously 

reviewed; or 

(3) Result in there no longer being 

reasonable assurance that adequate 

funds will be available for 

decommissioning. 

(g) In taking actions permitted under 

§ 53.1540 following submittal of the 

PSDAR, the licensee must notify the 

NRC in writing, and send a copy to the 

affected State(s), before performing any 

decommissioning activity inconsistent 

with, or making any significant 

schedule change from, those actions and 

schedules described in the PSDAR, 

including changes that increase the 

decommissioning cost by more than 20 

percent from the previously provided 

DCE. 

(h) Licensees may use 

decommissioning trust funds consistent 

with the limitations of § 53.1045(a). 

Licensees must report on the status of 

decommissioning trust funds consistent 

with the requirements of § 53.1060. 

(i) Licensees must submit an 

application for termination of license in 

accordance with § 53.1070. The 

application for termination of license 

must be accompanied or preceded by a 

license termination plan to be submitted 

for NRC approval. 

(1) The license termination plan must 

be a supplement to the Final Safety 

Analysis Report or equivalent and must 

be submitted at least 2 years before 

termination of the license date. 

(2) The license termination plan must 

include— 

(i) A site characterization; 

(ii) Identification of remaining 

dismantlement activities; 

(iii) Plans for site remediation; 

(iv) Detailed plans for the final 

radiation survey; 

(v) A description of the end use of the 

site, if restricted; 

(vi) An updated site-specific estimate 

of remaining decommissioning costs; 

(vii) A supplement to the 

environmental report, pursuant to 

§ 51.53 of this chapter, describing any 

new information or significant 

environmental change associated with 

the licensee’s proposed termination 

activities; and 

(viii) Identification of parts, if any, of 

the facility or site that were released for 

use before approval of the license 

termination plan. 

(3) Following receipt of the license 

termination plan, the NRC must make 

the license termination plan publicly 

available and publish notice of its 

availability for public comment in the 

Federal Register. The NRC must also 

schedule a public meeting readily 

accessible to individuals in the vicinity 

of the licensee’s facility upon receipt of 

the license termination plan. The NRC 

must publish a notice in the Federal 

Register and in a forum, such as local 

newspapers, that is readily accessible to 

individuals in the vicinity of the site, 

announcing the date, time, and location 

of the meeting, along with a brief 

description of the purpose of the 

meeting. 

(j) If the license termination plan 

demonstrates that the remainder of 


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decommissioning activities will be 

performed in accordance with the 

regulations in this chapter, will not be 

inimical to the common defense and 

security or to the health and safety of 

the public, and will not have a 

significant effect on the quality of the 

environment and after notice to 

interested persons, the Commission will 

approve the plan, by license 

amendment, subject to such conditions 

and limitations as it deems appropriate 

and necessary and authorize 

implementation of the license 

termination plan. 

(k) The Commission will terminate 

the license if it determines that— 

(1) The remaining dismantlement has 

been performed in accordance with the 

approved license termination plan, and 

(2) The final radiation survey and 

associated documentation, including an 

assessment of dose contributions 

associated with parts released for use 

before approval of the license 

termination plan, demonstrate that the 

facility and site have met the criteria for 

decommissioning in subpart E of 10 

CFR part 20. 

§ 53.1075 Program requirements during 

decommissioning. 

(a) Licensees that have submitted the 

certifications required under § 53.1070 

must maintain a decommissioning fire 

protection program to address the 

potential for fires that could cause the 

release or spread of radioactive 

materials. 

(1) The objectives of the 

decommissioning fire protection 

program are to 

(i) Reasonably prevent these fires from 

occurring; 

(ii) Rapidly detect, control, and 

extinguish those fires that do occur and 

that could result in a radiological 

hazard; and 

(iii) Ensure that the risk of fireinduced radiological hazards to the 

public, environment, and plant 

personnel is minimized. 

(2) The licensee must assess the 

decommissioning fire protection 

program on a regular basis. The licensee 

must revise the decommissioning fire 

protection program documentation as 

appropriate throughout the various 

stages of facility decommissioning. 

(3) The licensee may make changes to 

the decommissioning fire protection 

program without NRC approval if these 

changes do not reduce the effectiveness 

of fire protection for structures, systems, 

and components that could result in a 

radiological hazard, taking into account 

the decommissioning plant conditions 

and activities. 

(b) [Reserved] 

§ 53.1080 Release of part of a commercial 

nuclear plant or site for unrestricted use. 

(a) Prior written NRC approval is 

required to release part of a commercial 

nuclear plant or site for unrestricted use 

at any time before receiving approval of 

a license termination plan. Section 

53.1060 specifies recordkeeping 

requirements associated with partial 

release. Holders of an OL or COL under 

this part seeking NRC review and 

approval must— 

(1) Evaluate the effect of releasing the 

property to ensure that— 

(i) The dose to individual members of 

the public does not exceed the limits 

and standards of subpart D of 10 CFR 

part 20; 

(ii) There is no reduction in the 

effectiveness of emergency planning or 

physical security; 

(iii) Effluent releases remain within 

license conditions; 

(iv) The environmental monitoring 

program and offsite dose calculation 

manual are revised to account for the 

changes; 

(v) The siting criteria of subpart D of 

this part continue to be met; and 

(vi) All other applicable statutory and 

regulatory requirements continue to be 

met. 

(2) Perform a historical site 

assessment of the part of the commercial 

nuclear plant or site to be released; and 

(3) Perform surveys adequate to 

demonstrate compliance with the 

radiological criteria for unrestricted use 

specified in § 20.1402 of this chapter for 

impacted areas. 

(b) For release of non-impacted areas, 

the licensee may submit a written 

request for NRC review and approval of 

the release if a license amendment is not 

otherwise required. The request 

submittal must include— 

(1) The results of the evaluations 

performed in accordance with 

paragraphs (a)(1) and (a)(2) of this 

section; 

(2) A description of the part of the 

commercial nuclear plant or site to be 

released; 

(3) The schedule for release of the 

property; 

(4) The results of the evaluations 

performed in accordance with 

§ 53.1540; and 

(5) A discussion that provides the 

reasons for concluding that the 

environmental impacts associated with 

the licensee’s proposed release of the 

property will be bounded by 

appropriate previously issued 

environmental impact statements. 

(c) After receiving a request from the 

licensee for NRC approval of the release 

of a non-impacted area, the NRC must— 

(1) Determine whether the licensee 

has adequately evaluated the effect of 

releasing the property as required by 

paragraph (a)(1) of this section; 

(2) Determine whether the licensee’s 

classification of any release areas as 

non- impacted is adequately justified; 

and 

(3) If determining that the licensee’s 

submittal is adequate, inform the 

licensee in writing that the release is 

approved. 

(d) For release of impacted areas, the 

licensee must submit an application for 

amendment of its license for the release 

of the property. The application must 

include— 

(1) The information specified in 

paragraphs (b)(1) through (b)(3) of this 

section; 

(2) The methods used for and results 

obtained from the radiation surveys 

required to demonstrate compliance 

with the radiological criteria for 

unrestricted use specified in § 20.1402; 

and 

(3) A supplement to the 

environmental report, under § 51.53 of 

this chapter, describing any new 

information or significant 

environmental change associated with 

the licensee’s proposed release of the 

property. 

(e) After receiving a license 

amendment application from the 

licensee for the release of an impacted 

area, the NRC must— 

(1) Determine whether the licensee 

has adequately evaluated the effect of 

releasing the property as required by 

paragraph (a)(1) of this section; 

(2) Determine whether the licensee’s 

classification of any release areas as 

non-impacted is adequately justified; 

(3) Determine whether the licensee’s 

radiation survey for an impacted area is 

adequate; and 

(4) If determining that the licensee’s 

submittal is adequate, approve the 

licensee’s amendment application. 

(f) The NRC must publish notice 

receipt of the release approval request or 

license amendment application in the 

Federal Register and make the approval 

request or license amendment 

application available for public 

comment. Before acting on an approval 

request or license amendment 

application submitted in accordance 

with this section, the NRC must conduct 

a public meeting readily accessible to 

individuals in the vicinity of the 

licensee’s facility for the purpose of 

obtaining public comments on the 

proposed release of part of the 

commercial nuclear plant or site. The 

NRC must publish a document in the 

Federal Register and in a forum, such 

as local newspapers, which is readily 

accessible to individuals in the vicinity 

of the site, announcing the date, time, 


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and location of the meeting, along with 

a brief description of the purpose of the 

meeting. 

Subpart H—Licenses, Certifications, 

and Approvals 

§ 53.1100 Filing of application for licenses, 

certifications, or approvals; oath or 

affirmation. 

(a) Serving of applications. (1) Each 

filing of an application for a standard 

design approval, standard design 

certification, or license under this part, 

and any amendments to the 

applications, must be submitted to the 

U.S. Nuclear Regulatory Commission 

(NRC) under § 53.040, as applicable. 

(2) Each applicant for a construction 

permit (CP), early site permit, combined 

license (COL), or manufacturing license 

(ML) under this part must, upon 

notification by the presiding officer 

designated to conduct the public 

hearing required by the Atomic Energy 

Act of 1954, as amended, (the Act) 

update the application and serve the 

updated copies of the application or 

parts of it, eliminating all superseded 

information, together with an index of 

the updated application, as directed by 

presiding officer. Any subsequent 

amendment to the application must be 

served on those served copies of the 

application and must be submitted to 

the NRC as specified in § 53.040, as 

applicable. 

(3) The applicant must make a copy 

of the updated application available at 

the public hearing for the use of any 

other parties to the proceeding and must 

certify that the updated copies of the 

application contain the current contents 

of the application submitted in 

accordance with the requirements under 

this part. 

(4) At the time of filing an 

application, the Commission will make 

available at the NRC website, http://

www.nrc.gov, a copy of the application, 

subsequent amendments, and other 

records pertinent to the matter that is 

the subject of the application for public 

inspection and copying. 

(5) The serving of copies required by 

this section must not occur until the 

application has been docketed under 

§ 2.101(a) of this chapter. Copies must 

be submitted to the Commission, as 

specified in § 53.040, as applicable, to 

enable the Director, Office of Nuclear 

Reactor Regulation to determine 

whether the application is sufficiently 

complete to permit docketing. 

(b) Oath or affirmation. Each 

application for a standard design 

approval, standard design certification, 

or license, including, whenever 

appropriate, a CP or early site permit, or 

amendment of it, and each amendment 

of each application must be executed in 

a signed original by the applicant or 

duly authorized officer thereof under 

oath or affirmation. 

(c) [Reserved] 

(d) [Reserved] 

(e) Filing fees. Each application for a 

standard design approval, standard 

design certification, or commercial 

nuclear plant license under this part, 

including, whenever appropriate, a CP, 

COL, operating license (OL), ML, or 

early site permit, other than a license 

exempted from 10 CFR part 170, must 

be accompanied by the fee prescribed in 

10 CFR part 170. No fee will be required 

to accompany an application for 

renewal, amendment, or termination of 

a CP, OL, COL, or ML, except as 

provided in § 170.21 of this chapter. 

(f) Environmental report. An 

application for a CP, OL, early site 

permit, design certification, COL, or ML 

for a commercial nuclear plant must be 

accompanied by an environmental 

report required under subpart A of 10 

CFR part 51. 

§ 53.1101 Requirement for license. 

Except as provided in § 53.1120, no 

person within the United States may 

transfer or receive in interstate 

commerce, manufacture, produce, 

transfer, acquire, possess, or use any 

utilization facility except as authorized 

by a license issued by the Commission. 

§ 53.1103 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. 

§ 53.1106 Elimination of repetition. 

An applicant may incorporate by 

reference in its application information 

contained in previous applications, 

statements, or reports filed with the 

Commission, provided, however, that 

such references are clear and specific. 

§ 53.1109 Contents of applications; 

general information. 

Each application must include, unless 

otherwise indicated in this subpart— 

(a) Name of applicant; 

(b) Address of applicant; 

(c) Description of business or 

occupation of applicant; 

(d)(1) If applicant is an individual, the 

citizenship of applicant; 

(2) If applicant is a partnership, the 

name, citizenship and address of each 

partner and the principal location where 

the partnership does business; 

(3) If applicant is a corporation or an 

unincorporated association, the 

following information: 

(i) The State where it is incorporated 

or organized and the principal location 

where it does business; 

(ii) The names, addresses and 

citizenship of its directors and of its 

principal officers; and 

(iii) Whether it is owned, controlled, 

or dominated by an alien, a foreign 

corporation, or foreign government, and 

if so, give details; or 

(4) If the applicant is acting as agent 

or representative of another person in 

filing the application, identify the 

principal and furnish information 

required under this paragraph with 

respect to such principal; 

(e) The class and type of license 

applied for, the use to which the facility 

will be put, the period of time for which 

the license is sought, and a list of other 

licenses, except operator’s licenses, 

issued or applied for in connection with 

the proposed facility; 

(f) [Reserved] 

(g)(1) Except as provided in paragraph 

(g)(2) of this section, if the application 

is for an OL or COL for a commercial 

nuclear plant, or if the application is for 

an early site permit for a commercial 

nuclear plant and contains plans for 

coping with emergencies under 

§ 53.1146(b)(2)(ii), the applicant must 

submit the radiological emergency 

response plans of State, local, and 

participating Tribal governmental 

entities in the United States that are 

wholly or partially within the plume 

exposure pathway emergency planning 

zone (EPZ),1 and the plans of State 

governments wholly or partially within 

the ingestion pathway EPZ.2 If the 

application is for an early site permit 

that, under § 53.1146(b)(2)(i), proposes 

major features of the emergency plans 

describing the EPZs, then the 

descriptions of the EPZs must meet the 

requirements of this paragraph. 

Generally, the plume exposure pathway 

EPZ for a commercial nuclear plant 

must consist of an area about 10 miles 

(16 km) in radius and the ingestion 

pathway EPZ must consist of an area 

about 50 miles (80 km) in radius. The 

exact size and configuration of the EPZs 

surrounding a particular commercial 

nuclear plant must be determined in 

relation to the local emergency response 

needs and capabilities as they are 

affected by such conditions as 

demography, topography, land 

characteristics, access routes, and 

jurisdictional boundaries. The size of 

the EPZs also may be determined on a 

case-by-case basis for gas-cooled 

reactors and for reactors with an 

authorized power level less than 250 


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megawatt thermal. The plans for the 

ingestion pathway must focus on such 

actions as are appropriate to protect the 

food ingestion pathway. 

(2) Applicants for commercial nuclear 

plants consisting of either small 

modular reactors or non-light-water 

reactors complying with § 50.160 of this 

chapter who apply for a CP, an OL, a 

COL, or an early site permit under this 

part must submit as part of the 

application the analysis used to 

determine whether the criteria in 

§ 53.1109(g)(2)(i)(A) and (B) are met 

and, if they are met, the size of the 

plume exposure pathway EPZ. 

(i) The plume exposure pathway EPZ 

is the area within which: 

(A) Public dose, as defined in 

§ 20.1003 of this chapter, is projected to 

exceed 10 millisieverts (1 rem) total 

effective dose equivalent over 96 hours 

from the release of radioactive materials 

from the facility considering accident 

likelihood and source term, timing of 

the accident sequence, and meteorology; 

and 

(B) Pre-determined, prompt protective 

measures are necessary. 

(ii) If the application is for an OL or 

COL or if the application is for an early 

site permit and contains plans for 

coping with emergencies under 

§ 53.1146(b)(2)(ii), and if the plume 

exposure pathway EPZ extends beyond 

the site boundary: 

(A) The applicant must submit 

radiological emergency response plans 

of State, local, and participating Tribal 

governmental entities in the United 

States that are wholly or partially within 

the plume exposure pathway EPZ. 

(B) The exact configuration of the 

plume exposure pathway EPZ 

surrounding the facility shall be 

determined in relation to the local 

emergency response needs and 

capabilities as they are affected by such 

conditions as demography, topography, 

land characteristics, access routes, and 

jurisdictional boundaries. 

(iii) If the application is for an early 

site permit that, under § 53.1146(b)(2)(i), 

proposes major features of the 

emergency plans and describes the EPZ, 

and if the EPZ extends beyond the site 

boundary, then the exact configuration 

of the plume exposure pathway EPZ 

surrounding the facility must be 

determined in relation to the local 

emergency response needs and 

capabilities as they are affected by such 

conditions as demography, topography, 

land characteristics, access routes, and 

jurisdictional boundaries. 

(h) [Reserved] 

(i) A list of the names and addresses 

of such regulatory agencies as may have 

jurisdiction over the rates and services 

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. 

1EPZs are discussed in NUREG–0396, U.S. 

Environmental Protection Agency 520/1–78– 

016, ‘‘Planning Basis for the Development of 

State and Local Government Radiological 

Emergency Response Plans in Support of 

Light-Water Nuclear Power Plants,’’ 

December 1978. 2 If the State, local, and participating Tribal 

emergency response plans have been 

previously provided to the NRC for inclusion 

in the facility docket, the applicant need only 

provide the appropriate reference to meet 

this requirement. 

§ 53.1112 Environmental conditions. 

(a) Each CP, early site permit, and 

COL under this part may include 

conditions to address environmental 

issues during construction. These 

conditions are to be set out in an 

attachment to the license, which is 

incorporated in and made a part of the 

license. These conditions will be 

derived from information contained in 

the environmental report submitted 

pursuant to § 51.50 of this chapter, as 

analyzed and evaluated in the NRC 

record of decision and will identify the 

obligations of the licensee in the 

environmental area, including, as 

appropriate, requirements for reporting 

and keeping records of environmental 

data, and any conditions and 

monitoring requirement for the 

protection of the nonaquatic 

environment. 

(b) Each license authorizing operation 

of a commercial nuclear plant under 

this part, and each license for a 

commercial nuclear plant for which the 

certification of permanent cessation of 

operations required under § 53.1070 has 

been submitted may include conditions 

to address environmental issues during 

operation and decommissioning. These 

conditions are to be set out in an 

attachment to the license, which is 

incorporated in and made a part of the 

license. These conditions will be 

derived from information contained in 

the environmental report or the 

supplement to the environmental report 

submitted under §§ 51.50 and 51.53 of 

this chapter as analyzed and evaluated 

in the NRC record of decision, and will 

identify the obligations of the licensee 

in the environmental area, including, as 

appropriate, requirements for reporting 

and keeping records of environmental 

data and any conditions and monitoring 

requirement for the protection of the 

nonaquatic environment. 

§ 53.1115 Agreement limiting access to 

classified information. 

As part of its application and in any 

event before the receipt of Restricted 

Data or classified National Security 

Information or the issuance of a license 

or standard design approval under this 

part, or before the Commission has 

adopted a final standard design 

certification rule under this part, the 

applicant must agree in writing that it 

will not permit any individual to have 

access to or any facility to possess 

Restricted Data or classified National 

Security Information until the 

individual and/or facility has been 

approved for access under the 

provisions of 10 CFR parts 25 and/or 95. 

The agreement of the applicant becomes 

part of the license or standard design 

approval. 

§ 53.1118 Ineligibility of certain applicants. 

Any person who is a citizen, national, 

or agent of a foreign country, or any 

corporation, or other entity which the 

Commission knows or has reason to 

believe is owned, controlled, or 

dominated by an alien, a foreign 

corporation, or a foreign government, 

will be ineligible to apply for and obtain 

a license. 

§ 53.1120 Exceptions and exemptions 

from licensing requirements. 

Nothing in this part must be deemed 

to require a license for— 

(a) The manufacture, production, or 

acquisition by the Department of 

Defense of any utilization facility 

authorized pursuant to section 91 of the 

Act or the use of such facility by the 

Department of Defense or by a person 

under contract with and for the account 

of the Department of Defense; 

(b) Except to the extent that the 

Department of Energy facilities of the 

types subject to licensing pursuant to 

section 202 of the Energy 

Reorganization Act of 1974, as 

amended, are involved— 

(1)(i) The processing, fabrication or 

refining of special nuclear material 

(SNM) or the separation of SNM, or the 

separation of SNM from other 

substances by a prime contractor of the 

Department of Energy under a prime 

contract for— 


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(A) The performance of work for the 

Department of Energy at a United States 

government-owned or controlled site; 

(B) Research in, or development, 

manufacture, storage, testing or 

transportation of, atomic weapons or 

components thereof; or 

(C) The use or operation of a 

utilization facility in a United States 

owned vehicle or vessel; or 

(ii) The processing, fabrication or 

refining of SNM of the separation of 

SNM, or the separation of SNM from 

other substances by a prime contractor 

or subcontractor of the Commission or 

the Department of Energy under a prime 

contract or subcontract when the 

Commission determines that the 

exemption of the prime contractor or 

subcontractor is authorized by law; and 

that, under the terms of the contract or 

subcontract, there is adequate assurance 

that the work thereunder can be 

accomplished without undue risk to the 

public health and safety; or 

(2)(i) The construction or operation of 

a utilization facility for the Department 

of Energy at a United States 

government-owned or controlled site, 

including the transportation of the 

utilization facility to or from such site 

and the performance of contract services 

during temporary interruptions of such 

transportation; or the construction or 

operation of a utilization facility for the 

Department of Energy in the 

performance of research in, or 

development, manufacture, storage, 

testing, or transportation of, atomic 

weapons or components thereof; or the 

use or operation of a utilization facility 

for the Department of Energy in a 

United States government-owned 

vehicle or vessel; provided that such 

activities are conducted by a prime 

contractor of the Department of Energy 

under a prime contract with the 

Department of Energy; or 

(ii) The construction or operation of a 

utilization facility by a prime contractor 

or subcontractor of the Commission or 

the Department of Energy under his or 

her prime contract or subcontract when 

the Commission determines that the 

exemption of the prime contractor or 

subcontractor is authorized by law; and 

that, under the terms of the contract or 

subcontract, there is adequate assurance 

that the work thereunder can be 

accomplished without undue risk to the 

public health and safety; or 

(c) The transportation or possession of 

any utilization facility by a common or 

contract carrier or warehouse employee 

in the regular course of carriage for 

another or storage incident thereto. 

§ 53.1121 Public inspection of 

applications. 

Applications and documents 

submitted to the Commission in 

connection with applications may be 

made available for public inspection 

under the provisions of part 2 of this 

chapter. 

§ 53.1124 Relationship between sections. 

(a) Limited work authorization. An 

application for a limited work 

authorization (LWA) under this part 

may be submitted as part of an 

application for an early site permit, CP, 

or COL under this part as required in 

§ 53.1130(a)(2). 

(b) Early site permit. (1) A holder of 

an early site permit may request an 

LWA. 

(2) An application for a CP or COL 

under this part may, but need not, 

reference an early site permit. 

(c) Standard design approval. An 

application for a standard design 

approval under this part may, but need 

not, reference an OL or custom COL 

under this part that is essentially the 

same as the information supporting the 

standard design for which approval is 

being requested. 

(d) Standard design certification. An 

application for a standard design 

certification under this part may, but 

need not, reference an OL or custom 

COL under this part that is essentially 

the same as the standard design for 

which certification is being requested. 

(e) Manufacturing license. (1) A 

manufactured reactor manufactured 

under an ML issued under this part may 

only be transported to and installed at 

a site for which a COL under this part 

has been issued. 

(2) An ML applicant under this part 

may reference a standard design 

certification or a standard design 

approval under this part in its 

application. 

(f) Construction permit. An 

application for a CP may, but need not, 

reference a standard design certification 

or standard design approval issued 

under this part, respectively, and may 

also reference an early site permit 

issued under this part. In the absence of 

a demonstration that an entity other 

than the one originally sponsoring a 

standard design certification is qualified 

to supply a design, the Commission will 

entertain an application for a CP that 

references a standard design 

certification issued under this part only 

if the entity that sponsored the 

certification supplies the design for the 

applicant’s use. 

(g) Operating license. (1) An 

application for an OL under this part 

may, but need not, reference an early 

site permit, standard design 

certification, or standard design 

approval issued under this part. In the 

absence of a demonstration that an 

entity other than the one originally 

sponsoring a standard design 

certification is qualified to supply a 

design, the Commission will entertain 

an application for an OL that references 

a standard design certification issued 

under this part only if the entity that 

sponsored the certification supplies the 

design for the applicant’s use. 

(2) The holder of a CP must, at the 

time of submission of the Final Safety 

Analysis Report (FSAR), file an 

application for an OL. 

(h) Combined licenses. An application 

for a COL under this part may, but need 

not, reference an early site permit, 

standard design certification, standard 

design approval, or ML issued under 

this part. In the absence of a 

demonstration that an entity other than 

the one originally sponsoring and 

obtaining a standard design certification 

is qualified to supply a design, the 

Commission will entertain an 

application for a COL that references a 

standard design certification issued 

under this part only if the entity that 

sponsored the certification supplies the 

design for the applicant’s use. 

§ 53.1130 Limited work authorizations. 

(a) Request for limited work 

authorization. (1) Any person to whom 

the Commission may otherwise issue 

either a license or permit related to a 

commercial nuclear plant may request 

an LWA allowing that person to perform 

the driving of piles, subsurface 

preparation, placement of backfill, 

concrete, or permanent retaining walls 

within an excavation, and installation of 

the foundation, including placement of 

concrete, any of which are for a 

structure, system, or component (SSC) 

of the facility for which either a CP or 

COL is otherwise required under 

§ 53.610. 

(2) An application for an LWA may be 

submitted as part of a complete 

application for a CP or COL in 

accordance with § 2.101(a)(1) through 

(a)(5) of this chapter, or as a partial 

application in accordance with 

§ 2.101(a)(9) of this chapter. An 

application for an LWA by the holder of 

an early site permit must be submitted 

as a complete application in accordance 

with § 2.101(a)(1) through (a)(4) of this 

chapter. 

(3) The application must include— 

(i) A Safety Analysis Report required 

by §§ 53.1146, 53.1309 or 53.1416, as 

applicable, a description of the activities 

requested to be performed, and the 

design and construction information 


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otherwise required by the Commission’s 

rules and regulations to be submitted for 

a CP or COL under this part but limited 

to those portions of the facility that are 

within the scope of the LWA. The Safety 

Analysis Report must demonstrate that 

activities conducted under the LWA 

will be conducted in compliance with 

the technically relevant Commission 

requirements in 10 CFR chapter I 

applicable to the design of those 

portions of the facility within the scope 

of the LWA; 

(ii) An environmental report in 

accordance with § 51.49 of this chapter; 

and 

(iii) A plan for redress of activities 

performed under the LWA, should 

limited work activities be terminated by 

the holder, or the LWA be revoked by 

the NRC or upon effectiveness of the 

Commission’s final decision denying 

the associated CP or COL application, as 

applicable. 

(b) Issuance of limited work 

authorization. (1) The Director, Office of 

Nuclear Reactor Regulation may issue 

an LWA only after— 

(i) The NRC staff issues the final 

environmental impact statement for the 

LWA under subpart A of part 51 of this 

chapter; 

(ii) The presiding officer makes the 

finding in §§ 51.105(c) or 51.107(d) of 

this chapter, as applicable; 

(iii) The Director determines that the 

applicable standards and requirements 

of the Act, and the Commission’s 

regulations applicable to the activities to 

be conducted under the LWA, have 

been met, the applicant is technically 

qualified to engage in the activities 

authorized, and that issuance of the 

LWA will provide reasonable assurance 

of adequate protection to public health 

and safety and will not be inimical to 

the common defense and security; and 

(iv) The presiding officer finds that 

there are no unresolved safety issues 

relating to the activities to be conducted 

under the LWA that would constitute 

good cause for withholding the 

authorization. 

(2) Each LWA will specify the 

activities that the holder is authorized to 

perform. 

(c) Effect of limited work 

authorization. Any activities 

undertaken under an LWA are entirely 

at the risk of the applicant and, except 

as to the matters determined under 

paragraph (b)(1) of this section, the 

issuance of the LWA has no bearing on 

the issuance of a CP or COL with respect 

to the requirements of the Act and rules, 

regulations, or orders issued under the 

Act. The environmental impact 

statement for a CP or COL application 

for which an LWA was previously 

issued will not address, and the 

presiding officer will not consider, the 

sunk costs of the holder of the LWA in 

determining the proposed action (i.e., 

issuance of the CP or COL). 

(d) Implementation of redress plan. If 

construction is terminated by the 

holder, the underlying application is 

withdrawn by the applicant or denied 

by the NRC, or the LWA is revoked by 

the NRC, then the holder must begin 

implementation of the redress plan in a 

reasonable time. The holder must also 

complete the redress of the site no later 

than 18 months after termination of 

construction, revocation of the LWA, or 

upon effectiveness of the Commission’s 

final decision denying the associated CP 

application or the associated COL 

application, as applicable. 

§ 53.1140 Early site permits. 

Sections 53.1140 through 53.1188 set 

out the requirements and procedures 

applicable to Commission issuance of 

an early site permit under this part for 

approval of a site for a commercial 

nuclear plant separate from the filing of 

an application for a CP or COL for the 

facility. 

§ 53.1143 Filing of applications. 

Any person who may apply for a CP 

or for a COL under this part, may file 

an application for an early site permit 

with the Director, Office of Nuclear 

Reactor Regulation. An application for 

an early site permit may be filed 

notwithstanding the fact that an 

application for a CP or a COL has not 

been filed in connection with the site 

for which a permit is sought. 

§ 53.1144 Contents of applications for 

early site permits; general information. 

The application must contain all of 

the information required by § 53.1109(a) 

through (d) and (j). 

§ 53.1146 Contents of applications for 

early site permits; technical information. 

(a) The application must contain— 

(1) A Site Safety Analysis Report that 

must include the following: 

(i) The specific number, type, and 

thermal power level of the facilities, or 

range of possible facilities, for which the 

site may be used; 

(ii) The anticipated maximum levels 

of radiological and thermal effluents 

each facility will produce; 

(iii) The type of cooling systems, 

including intakes and outflows, where 

appropriate, that may be associated with 

each facility; 

(iv) The boundaries of the site; 

(v) The proposed general location of 

each facility on the site; 

(vi) The external hazards and site 

characteristics required by this part; 

(vii) The location and description of 

any nearby industrial, military, or 

transportation facilities and routes; 

(viii) The existing and projected 

future population profile of the area 

surrounding the site; 

(ix) A description and assessment of 

the site on which a facility is to be 

located. The assessment must address 

the requirements of subpart D of this 

part; 

(x) Information demonstrating that 

site characteristics are such that 

adequate security plans and measures 

can be developed; and 

(xi) A description of the quality 

assurance program (QAP) required by 

appendix B to part 50 of this chapter 

applied to site-related activities for the 

future design, fabrication, construction, 

and testing of the SSCs of a facility or 

facilities that may be constructed on the 

site. 

(2) A complete environmental report 

as required by § 51.50(b) of this chapter. 

(b)(1) The Site Safety Analysis Report 

must identify physical characteristics of 

the proposed site, such as egress 

limitations from the area surrounding 

the site, that could pose a significant 

impediment to the development of 

emergency plans. If physical 

characteristics are identified that could 

pose a significant impediment to the 

development of emergency plans, the 

application must identify measures that 

would, when implemented, mitigate or 

eliminate the significant impediment. 

(2) The Site Safety Analysis Report 

may also— 

(i) Propose major features of the 

emergency plans, under either § 50.160 

or the requirements in appendix E to 

part 50 and § 50.47(b) of this chapter, as 

applicable, such as the exact size and 

configuration of the EPZs, for review 

and approval by the NRC, in 

consultation with the Federal 

Emergency Management Agency 

(FEMA), as applicable, in the absence of 

complete and integrated emergency 

plans; or 

(ii) Propose complete and integrated 

emergency plans for review and 

approval by the NRC, in consultation 

with FEMA, as applicable, in 

accordance with either § 50.160 or the 

requirements in appendix E to part 50 

and § 50.47(b) of this chapter. To the 

extent approval of emergency plans is 

sought, the application must contain the 

information required by § 53.1109(g). 

(3) Emergency plans submitted under 

paragraph (b)(2)(ii) of this section must 

include the proposed inspections, tests, 

and analyses that the holder of a COL 

referencing the early site permit must 

perform, and the acceptance criteria that 

are necessary and sufficient to provide 


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reasonable assurance that, if the 

inspections, tests, and analyses are 

performed and the acceptance criteria 

met, the facility has been constructed 

and will be operated in conformity with 

the emergency plans, the provisions of 

the Act, and the Commission’s rules and 

regulations. Major features of an 

emergency plan submitted under 

paragraph (b)(2)(i) of this section may 

include proposed inspections, tests, 

analyses, and acceptance criteria 

(ITAAC). 

(4) Under paragraphs (b)(1) and 

(b)(2)(i) of this section, the Site Safety 

Analysis Report must include, where 

appropriate, a description of contacts 

and arrangements made with Federal, 

State, participating Tribal, and local 

governmental agencies with emergency 

planning responsibilities. The Site 

Safety Analysis Report must contain any 

certifications that have been obtained. If 

these certifications, where appropriate, 

cannot be obtained, the Site Safety 

Analysis Report must contain 

information, including a utility plan, 

sufficient to show that the proposed 

plans provide reasonable assurance that 

adequate protective measures can and 

will be taken in the event of a 

radiological emergency at the site. 

Under the option set forth in paragraph 

(b)(2)(ii) of this section, the applicant 

must make good faith efforts, where 

appropriate, to obtain from the same 

governmental agencies certifications 

that— 

(i) The proposed emergency plans are 

practicable; 

(ii) These agencies are committed to 

participating in any further 

development of the plans, including any 

required field demonstrations; and 

(iii) That these agencies are 

committed to executing their 

responsibilities under the plans in the 

event of an emergency. 

(c) An applicant may request that an 

LWA under § 53.1130 be issued in 

conjunction with the early site permit. 

The application must include the 

information otherwise required by 

§ 53.1130. 

(d) Each applicant for an early site 

permit under this part must protect 

safeguards information against 

unauthorized disclosure in accordance 

with the requirements in §§ 73.21 and 

73.22 of this chapter, as applicable. 

§ 53.1149 Review of applications. 

(a) Standards for review of 

applications. Applications filed under 

this part will be reviewed according to 

the applicable standards set out in this 

part. In addition, the Commission must 

prepare an environmental impact 

statement during review of the 

application, under the applicable 

provisions of 10 CFR part 51. The 

Commission must determine, after 

consultation with FEMA, as applicable, 

whether the information required of the 

applicant by § 53.1146(b)(1) shows that 

there is no significant impediment to 

the development of emergency plans 

that cannot be mitigated or eliminated 

by measures proposed by the applicant, 

whether any major features of 

emergency plans submitted by the 

applicant under § 53.1146(b)(2)(i) are 

acceptable under either § 50.160 or 

appendix E to part 50 and § 50.47(b) of 

this chapter, and whether any 

emergency plans submitted by the 

applicant under § 53.1146(b)(2)(ii) 

provide reasonable assurance that 

adequate protective measures can and 

will be taken in the event of a 

radiological emergency. 

(b) Administrative review of 

applications; hearings. An early site 

permit application is subject to all 

procedural requirements in 10 CFR part 

2, including the requirements for 

docketing in § 2.101(a)(1) through (4) of 

this chapter, and the requirements for 

issuance of a notice of hearing in 

§ 2.104(a) and (d) of this chapter, 

provided that the designated sections 

may not be construed to require that the 

environmental report, or draft or final 

environmental impact statement 

includes an assessment of the benefits of 

construction and operation of the 

reactor or reactors, or an analysis of 

alternative energy sources. The 

presiding officer in an early site permit 

hearing must not admit contentions 

proffered by any party concerning an 

assessment of the benefits of 

construction and operation of the 

reactor or reactors, or an analysis of 

alternative energy sources if those issues 

were not addressed by the applicant in 

the early site permit application. All 

hearings conducted on applications for 

early site permits filed under this part 

are governed by the procedures 

contained in subparts C, G, L, and N of 

10 CFR part 2, as applicable. 

§ 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. 

§ 53.1158 Issuance of early site permit. 

(a) After conducting a hearing under 

§ 53.1149(b) and receiving the report to 

be submitted by the ACRS under 

§ 53.1155, the Commission may issue an 

early site permit, in the form the 

Commission deems appropriate, if the 

Commission finds that— 

(1) An application for an early site 

permit demonstrates compliance with 

the applicable standards and 

requirements of the Act and the 

Commission’s regulations; 

(2) Notifications, if any, to other 

agencies or bodies have been duly 

made; 

(3) There is reasonable assurance that 

the site is in conformity with the 

provisions of the Act and the 

Commission’s regulations; 

(4) The applicant is technically 

qualified to engage in any activities 

authorized; 

(5) The proposed ITAAC, including 

any on emergency planning, are 

necessary and sufficient, within the 

scope of the early site permit, to provide 

reasonable assurance that the facility 

has been constructed and will be 

operated in conformity with the license, 

the provisions of the Act, and the 

Commission’s regulations; 

(6) Issuance of the permit will not be 

inimical to the common defense and 

security or to the health and safety of 

the public; 

(7) Any significant adverse 

environmental impact resulting from 

activities requested under § 53.1146(c) 

can be redressed; and 

(8) The findings required by subpart 

A of 10 CFR part 51 have been made. 

(b) The early site permit must specify 

the site characteristics, design 

parameters, and terms and conditions of 

the early site permit the Commission 

deems appropriate. Before issuance of 

either a CP or COL referencing an early 

site permit, the Commission must find 

that any relevant terms and conditions 

of the early site permit have been met. 

Any terms or conditions of the early site 

permit that could not be met by the time 

of issuance of the CP or COL, must be 

set forth as terms or conditions of the CP 

or COL. 

(c) The early site permit must specify 

those § 53.1130(b) activities requested 

under § 53.1146(c) that the permit 

holder is authorized to perform. 

§ 53.1161 Extent of activities permitted. 

If the activities authorized by 

§ 53.1158(c) are performed and the site 

is not referenced in an application for a 

CP or a COL issued under this part 

while the permit remains valid, then the 

early site permit remains in effect solely 

for the purpose of site redress, and the 

holder of the permit must redress the 

site under the terms of the site redress 

plan required by § 53.1146(c). If, before 

redress is complete, a use not envisaged 

in the redress plan is found for the site 

or parts thereof, the holder of the permit 


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must carry out the redress plan to the 

greatest extent possible consistent with 

the alternate use. 

§ 53.1164 Duration of permit. 

(a) Except as provided in paragraph 

(b) of this section, an early site permit 

issued under this subpart may be valid 

for not less than 10, nor more than 20 

years from the date of issuance. 

(b) An early site permit continues to 

be valid beyond the date of expiration 

in any proceeding on a CP application 

or a COL application that references the 

early site permit and is docketed before 

the date of expiration of the early site 

permit, or, if a timely application for 

renewal of the permit has been 

docketed, before the Commission has 

determined whether to renew the 

permit. 

(c) An applicant for a CP or COL may, 

at its own risk, reference in its 

application a site for which an early site 

permit application has been docketed 

but not granted. 

(d) Upon issuance of a CP or COL, a 

referenced early site permit is 

subsumed, to the extent referenced, into 

the CP or COL. 

§ 53.1167 Limited work authorization after 

issuance of early site permit. 

A holder of an early site permit may 

request an LWA under § 53.1130. 

§ 53.1170 Transfer of early site permit. 

An application to transfer an early site 

permit will be processed under 

§ 53.1570. 

§ 53.1173 Application for renewal. 

(a) Not less than 12, nor more than 36 

months before the expiration date stated 

in the early site permit, or any later 

renewal period, the permit holder may 

apply for a renewal of the permit. An 

application for renewal must contain all 

information necessary to bring up to 

date the information and data contained 

in the previous application. 

(b) Any person whose interests may 

be affected by renewal of the permit 

may request a hearing on the 

application for renewal. The request for 

a hearing must comply with § 2.309 of 

this chapter. If a hearing is granted, 

notice of the hearing will be published 

under § 2.309 of this chapter. 

(c) An early site permit, either original 

or renewed, for which a timely 

application for renewal has been filed, 

remains in effect until the Commission 

has determined whether to renew the 

permit. If the permit is not renewed, it 

continues to be valid in certain 

proceedings in accordance with the 

provisions of § 53.1164(b). 

(d) The Commission must refer a copy 

of the application for renewal to the 

ACRS. The ACRS must report on those 

portions of the application which 

concern safety and must apply the 

criteria set forth in § 53.1176. 

§ 53.1176 Criteria for renewal. 

(a) The Commission must grant the 

renewal if it determines that— 

(1) The site complies with the Act, the 

Commission’s regulations, and orders 

applicable and in effect at the time the 

site permit was originally issued; and 

(2) Any new requirements the 

Commission may wish to impose— 

(i) Are necessary for adequate 

protection to public health and safety or 

common defense and security; 

(ii) Are necessary for compliance with 

the Commission’s regulations, and 

orders applicable and in effect at the 

time the site permit was originally 

issued; or 

(iii) Would provide a substantial 

increase in overall protection of the 

public health and safety or the common 

defense and security to be derived from 

the new requirements, and the direct 

and indirect costs of implementation of 

those requirements are justified in view 

of this increased protection. 

(b) A denial of renewal under the 

provisions of § 53.1176(a) does not bar 

the permit holder or another applicant 

from filing a new application for the site 

which proposes changes to the site or 

the way that it is used to correct the 

deficiencies cited in the denial of the 

renewal. 

§ 53.1179 Duration of renewal. 

Each renewal of an early site permit 

may be for not less than 10, nor more 

than 20 years, plus any remaining years 

on the early site permit then in effect 

before renewal. 

§ 53.1182 Use of site for other purposes. 

A site for which an early site permit 

has been issued under this part may be 

used for purposes other than those 

described in the permit, including the 

location of other types of energy 

facilities. The permit holder must 

inform the Director, Office of Nuclear 

Reactor Regulation (Director), of any 

significant uses for the site which have 

not been approved in the early site 

permit. The information about the 

activities must be given to the Director 

at least 30 days in advance of any actual 

construction or site modification for the 

activities. The information provided 

could be the basis for imposing new 

requirements on the permit, under the 

provisions of § 53.1188. If the permit 

holder informs the Director that the 

holder no longer intends to use the site 

for a commercial nuclear plant, the 

Director may terminate the permit. 

§ 53.1188 Finality of early site permit 

determinations. 

(a) Commission finality. (1) While an 

early site permit is in effect under 

§ 53.1164 or § 53.1179, the Commission 

may not change or impose new site 

characteristics, design parameters, or 

terms and conditions, including 

emergency planning requirements, on 

the early site permit unless the 

Commission— 

(i) Determines that a modification is 

necessary to bring the permit or the site 

into compliance with the Commission’s 

regulations and orders applicable and in 

effect at the time the permit was issued; 

(ii) Determines the modification is 

necessary to assure adequate protection 

of the public health and safety or the 

common defense and security; 

(iii) Determines that a modification is 

necessary based on an update under 

paragraph (b) of this section; or 

(iv) Issues a variance requested under 

paragraph (d) of this section. 

(2) In making the findings required for 

issuance of a CP, COL, or OL, or the 

findings required by § 53.1452(g), or in 

any enforcement hearing other than one 

initiated by the Commission under 

paragraph (a)(1) of this section, if the 

application for the CP, COL, or OL 

references an early site permit, the 

Commission must treat as resolved 

those matters resolved in the proceeding 

on the application for issuance or 

renewal of the early site permit, except 

as provided for in paragraphs (b), (c), 

and (d) of this section. 

(i) If the Commission grants a CP 

application that references an early site 

permit and an application for an OL 

references the CP, the Commission must 

treat as resolved those matters resolved 

in the proceeding for the issuance or 

renewal of the early site permit, except 

as provided for in paragraphs (b), (c), 

and (d) of this section. 

(ii) If the early site permit approved 

an emergency plan (or major features 

thereof) that is in use by a licensee of 

a commercial nuclear plant, the 

Commission must treat as resolved 

changes to the early site permit 

emergency plan (or major features 

thereof) that are identical to changes 

made to the licensee’s emergency plans 

under § 53.1565 occurring after issuance 

of the early site permit. 

(iii) If the early site permit approved 

an emergency plan (or major features 

thereof) that is not in use by a licensee 

of a commercial nuclear plant, the 

Commission must treat as resolved 

changes that are equivalent to those that 

could be made under § 53.1565 without 

prior NRC approval had the emergency 

plan been in use by a licensee. 


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(b) Updating of early site permitemergency preparedness. An applicant 

for a CP, OL, or COL who has filed an 

application referencing an early site 

permit issued under this subpart must 

update the emergency preparedness 

information that was provided under 

§ 53.1146(b) and discuss whether the 

updated information materially changes 

the bases for compliance with 

applicable NRC requirements. 

(c) Hearings and petitions. (1) In any 

proceeding for the issuance of a CP, OL, 

or COL referencing an early site permit, 

contentions on the following matters 

may be litigated in the same manner as 

other issues material to the proceeding: 

(i) The nuclear reactor proposed to be 

built does not fit within one or more of 

the site characteristics or design 

parameters included in the early site 

permit; 

(ii) One or more of the terms and 

conditions of the early site permit have 

not been met; 

(iii) A variance requested under 

paragraph (d) of this section is 

unwarranted or should be modified; 

(iv) New or additional information is 

provided in the application that 

substantially alters the bases for a 

previous NRC conclusion or constitutes 

a sufficient basis for the Commission to 

modify or impose new terms and 

conditions related to emergency 

preparedness; or 

(v) Any significant environmental 

issue that was not resolved in the early 

site permit proceeding, or any issue 

involving the impacts of construction 

and operation of the facility that was 

resolved in the early site permit 

proceeding for which significant new 

information has been identified. 

(2) Any person may file a petition 

requesting that the site characteristics, 

design parameters, or terms and 

conditions of the early site permit be 

modified, or that the permit be 

suspended or revoked. The petition will 

be considered under § 2.206 of this 

chapter. Before construction 

commences, the Commission must 

consider the petition and determine 

whether any immediate action is 

required. If the petition is granted, an 

appropriate order will be issued. 

Construction under the CP or COL will 

not be affected by the granting of the 

petition unless the order is made 

immediately effective. Any change 

required by the Commission in response 

to the petition must demonstrate 

compliance with the requirements of 

paragraph (a)(1) of this section. 

(d) Variances. An applicant for a CP, 

OL, or COL referencing an early site 

permit may include in its application a 

request for a variance from one or more 

site characteristics, design parameters, 

or terms and conditions of the early site 

permit, or from the Site Safety Analysis 

Report. In determining whether to grant 

the variance, the Commission must 

apply the same technically relevant 

criteria applicable to the application for 

the original or renewed early site 

permit. Once a CP or COL referencing 

an early site permit is issued, variances 

from the early site permit will not be 

granted for that CP or COL. 

(e) Early site permit amendment. The 

holder of an early site permit may not 

make changes to the early site permit or 

the Site Safety Analysis Report without 

prior Commission approval. The request 

for a change to the early site permit 

must be in the form of an application for 

a license amendment and must 

demonstrate compliance with the 

requirements of §§ 53.1510 and 53.1520. 

§ 53.1200 Standard design approvals. 

Sections 53.1200 through 53.1221 set 

out procedures for the filing, NRC staff 

review, and referral to the ACRS of 

standard designs, or major portions 

thereof, for a commercial nuclear plant 

under this part. 

§ 53.1203 Filing of applications. 

Any person may submit a proposed 

standard design for a commercial 

nuclear plant to the NRC staff for its 

review. The submittal may consist of 

either the final design for the entire 

facility or the final design for major 

portions thereof. 

§ 53.1206 Contents of applications for 

standard design approvals; general 

information. 

The application must contain all of 

the information required by § 53.1109(a) 

through (c) and (j). 

§ 53.1209 Contents of applications for 

standard design approvals; technical 

information. 

(a) Major portion of a standard design. 

If the applicant seeks review of a major 

portion of a standard design, the 

application need only contain the 

information required by this section to 

the extent the requirements are 

applicable to the major portion of the 

standard design for which NRC staff 

approval is sought. If an applicant seeks 

approval of a major portion of the 

design, the scope of the application for 

which approval is sought must include 

all functional design criteria necessary 

to demonstrate compliance with the 

safety criteria in §§ 53.210, 53.220 and 

53.450(e), as applicable, for the major 

portion of the standard design for which 

NRC staff approval is sought. Such 

applicants must identify conditions 

related to interfaces with systems 

outside the scope of the major portion 

of the standard design for which NRC 

staff approval is sought, and functional 

or physical boundary conditions 

between the major portion of the 

standard design for which NRC staff 

approval is sought and the remainder of 

the standard design. These conditions 

must be demonstrated when the 

standard design approval is 

incorporated into a subsequent CP, 

design certification, ML, or COL 

application. 

(b) Final Safety Analysis Report. The 

application must contain an FSAR that 

describes the facility and the limits on 

its operation, presents a safety analysis 

of the SSCs and of the facility, or major 

portions thereof, for which the applicant 

seeks design approval, and must include 

the following information: 

(1) Site Parameters. The site 

parameters postulated for the design 

under this part, including the designbasis external hazard levels for the 

relevant external hazards, and an 

analysis and evaluation of the design in 

terms of those site parameters. 

(2) Design information. Except as 

specified in this paragraph, an 

application for a standard design 

approval for a commercial nuclear plant 

must include the design information 

equivalent to that required for a 

standard design certification under 

§ 53.1239(a)(2) through (27) for those 

portions of a commercial nuclear plant 

included in the standard design 

approval. 

§ 53.1210 Contents of applications for 

standard design approvals; other 

application content. 

(a) In addition to the FSAR, the 

application must also include the 

following: 

(1) Availability Controls (if not 

included in the FSAR). A description of 

the controls on plant operations, 

including availability controls, to 

provide reasonable confidence that the 

configurations and special treatments 

for safety-related (SR) SSCs and nonsafety-related but safety-significant 

(NSRSS) SSCs provide the capabilities 

and reliabilities required to demonstrate 

compliance with the safety criteria of 

§ 53.220. 

(2) Safeguards Information. A 

description of the program to protect 

Safeguards Information against 

unauthorized disclosure in accordance 

with the requirements in §§ 73.21 and 

73.22 of this chapter, as applicable. 

(b) If there are SSCs of the plant 

which required research and 

development to confirm the adequacy of 

their design, provide a report in the 

application which documents the 


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resolution of any safety questions 

associated with such SSCs. 

(c) A description of how the 

performance of each design feature has 

been demonstrated capable of fulfilling 

functional design criteria considering 

interdependent effects through either 

analysis, appropriate test programs, 

prototype testing, operating experience, 

or a combination thereof, in accordance 

with § 53.440(a). 

§ 53.1212 Standards for review of 

applications. 

Applications filed under this part will 

be reviewed under the standards set out 

in 10 CFR parts 20, 53, and 73. 

§ 53.1215 Referral to the Advisory 

Committee on Reactor Safeguards. 

The Commission must refer a copy of 

the application to the ACRS. The ACRS 

must report on those portions of the 

application which concern safety. 

§ 53.1218 Staff approval of design. 

(a) Upon completion of its review of 

a submittal under §§ 53.1200 through 

53.1221 and receipt of a report by the 

ACRS under § 53.1215, the NRC staff 

must publish a determination in the 

Federal Register as to whether or not 

the design is acceptable, subject to 

appropriate terms and conditions, and 

make an analysis of the design in the 

form of a report available at the NRC 

website, https://www.nrc.gov. 

(b) A standard design approval issued 

under this section is valid for 15 years 

from the date of issuance and may not 

be renewed. A design approval 

continues to be valid beyond the date of 

expiration in any proceeding on an 

application for a CP, OL, COL, or ML 

under this part that references the 

design approval and is docketed before 

the date of expiration of the design 

approval. 

§ 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 § 53.1580 and must be approved by 

the Executive Director for Operations or 

authorized designee before issuance of 

the request. 

(d) The Commission will require, 

before granting a CP, COL, OL, or ML 

that references a standard design 

approval, that engineering documents, 

such as analyses, drawings, 

procurement specifications, or 

construction and installation 

specifications, be completed and 

available for audit if the more detailed 

information is necessary for the 

Commission to verify the information in 

the application and make its safety 

determination, including the 

determination that the application is 

consistent with the design approval 

information. This information may be 

acquired by appropriate arrangements 

with the design approval applicant. 

§ 53.1230 Standard design certifications. 

Sections 53.1230 through 53.1263 set 

forth the requirements and procedures 

applicable to the Commission’s issuance 

of rules granting standard design 

certifications for commercial nuclear 

plants under this part separate from the 

filing of an application for a CP or COL 

for such a facility. 

§ 53.1233 Filing of applications. 

(a) An application for design 

certification may be filed 

notwithstanding the fact that an 

application for a CP, COL, or ML for 

such a facility has not been filed. 

(b) The application must comply with 

the applicable filing requirements of 

§ 53.040 and §§ 2.811 through 2.819 of 

this chapter. 

§ 53.1236 Contents of applications for 

standard design certifications; general 

information. 

The application must contain all of 

the information required by § 53.1109(a) 

through (c) and (j). 

§ 53.1239 Contents of applications for 

standard design certifications; technical 

information. 

The application must contain a level 

of design information sufficient to 

enable the Commission to judge the 

applicant’s proposed means of assuring 

that construction conforms to the design 

and to reach a final conclusion on all 

safety questions associated with the 

design before the certification is 

granted. The information submitted for 

a design certification must include 

performance requirements and design 

information sufficiently detailed to 

permit the preparation of acceptance 

and inspection requirements by the 

NRC. The Commission will require, 

before design certification, that 

information normally contained in 

engineering documents, such as 

analyses, drawings, procurement 

specifications, or construction and 

installation specifications, be completed 

and available for audit if the more 

detailed information is necessary for the 

Commission to verify the information in 

the application and make its safety 

determination. 

(a) Final Safety Analysis Report. The 

application must contain an FSAR that 

describes the facility and the limits on 

its operation, and presents a safety 

analysis of the SSCs, and must include 

the following information: 

(1) Site Parameters. The site 

parameters postulated for the design 

under this part, including the designbasis external hazard levels for the 

relevant external hazards, and an 

analysis and evaluation of the design in 

terms of those site parameters. 

(2) Plant Description and Safety 

Functions—(i) General Plant 

Description. A general description of the 

commercial nuclear plant including 

reactor type, the intended use of the 

reactor, nuclear design (e.g., neutron 

spectrum, reactor control, multi-unit 

reactor control), overall layout of the 

plant including significant plant 

features and SSCs, maximum power 

level and the nature and inventory of 

radioactive materials. 

(ii) Safety functions. A description of 

the primary and additional safety 

functions required under § 53.230 and a 

summary of how each safety function is 

satisfied. 

(3) Design Features and functional 

design criteria—licensing-basis events. 

(i) A description of the design features 

required by § 53.400 and the functional 

design criteria required by §§ 53.410 

and 53.420 that, when combined with 

corresponding human actions and 

programmatic controls, demonstrate that 

the plant will demonstrate compliance 

with the safety criteria defined in 

§ 53.210 and established in accordance 

with § 53.220, or more restrictive 

alternative criteria adopted under 

§ 53.470, during licensing-basis events 

(LBEs). 

(ii) A description of how design 

features demonstrate compliance with 


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the requirements of § 53.440(a) through 

(i) and (k) through (m). 

(4) Design Features Supporting 

Normal Operations. A description of the 

design features required by § 53.425 to 

support the holder of an OL or COL 

complying with § 53.260 during normal 

operations. 

(5) Design Features and Functional 

Design Criteria—aircraft impact. A 

description of the design features and 

functional design criteria required to 

demonstrate compliance with the 

requirements of § 53.440(j) for 

addressing the impact of a large, 

commercial aircraft. 

(6) Earthquake engineering. The 

information necessary to demonstrate 

that the commercial nuclear plant 

complies with the earthquake 

engineering criteria in § 53.480. 

(7) Programmatic Controls and 

Interfaces. (i) A description of the 

corresponding programmatic controls 

and interfaces necessary to achieve and 

maintain the reliability and capability of 

SSCs relied upon to demonstrate 

compliance with the functional design 

criteria required by §§ 53.410 and 

53.420 and the safety criteria in 

§§ 53.210 and 53.220, or more restrictive 

alternative criteria adopted under 

§ 53.470, and necessary to maintain 

consistency with analyses required by 

§ 53.450. 

(ii) For an application for a multi-unit 

commercial nuclear plant, the 

programmatic controls and interfaces 

must also be described for different 

modular configurations, as required by 

§ 53.440(i), including any restrictions 

that will be necessary during the 

construction and startup of any given 

unit to ensure the safe operation of the 

overall commercial nuclear plant to be 

licensed under this part. 

(8) Programmatic Controls for Normal 

Operations. A description of how 

programmatic controls, including 

monitoring programs, would provide 

assurance that design features and 

procedures will enable the holder of an 

OL or COL to comply with § 53.260. 

(9) Design Features Supporting the 

Protection of Plant Workers. A 

description of the design features 

required by § 53.430 to support the 

holder of an OL or COL complying with 

§ 53.270. 

(10) Programmatic Controls for 

Protection of Plant Workers. A 

description of how programmatic 

controls, including monitoring 

programs, would provide assurance that 

design features and procedures will 

enable the holder of an OL or COL to 

comply with § 53.270. 

(11) Codes and Standards. A 

description of generally accepted 

consensus codes and standards used to 

design the design features, as required 

by § 53.440(b). 

(12) Materials. A description of the 

materials used for SR and NSRSS SSCs 

and a description of the qualification of 

these materials for their service 

conditions over the plant lifetime, as 

required by § 53.440(c). 

(13) Integrity Assessment Program. A 

description of a design integrity 

assessment program that addresses the 

elements described in § 53.440(d). 

(14) Safety and Security. Confirmation 

that safety and security were considered 

together in the design process, as 

required by § 53.440(f). 

(15) Criticality. Information 

demonstrating how the applicant will 

comply with requirements for criticality 

accidents in § 53.440(m). 

(16) Multi-unit Plants. For an 

application for standard design 

certification of a multi-unit commercial 

nuclear plant, the possible operating 

configurations of the reactor units, 

including common systems, interface 

requirements, and system interactions, 

as required by § 53.440(i). 

(17) SSC Classification. (i) The 

classification of SSCs according to their 

safety significance under § 53.460(a). 

(ii) For SR and NSRSS SSCs, the 

conditions under which they must 

perform the safety functions required by 

§ 53.230, including environmental 

conditions. 

(18) Probabilistic Risk Assessment. A 

description of the probabilistic risk 

assessment (PRA) required by 

§ 53.450(a) and its results. 

(19) Analyses. A description of the 

analyses performed under § 53.450(b) 

through (g) that includes the following 

information: 

(i) A description of the analysis of 

LBEs and its results, as described in 

§ 53.240. This analysis description 

must— 

(A) Address the elements in 

§ 53.450(e) and (f); and 

(B) Under § 53.460(c)— 

(1) Describe any human actions that 

are necessary to prevent or mitigate 

LBEs; 

(2) Describe how those human actions 

are capable of being reliably performed 

under the postulated environmental 

conditions present; and 

(3) Describe how those human actions 

would be addressed by programs 

established under subpart F of this part. 

(ii)(A) A description of how SSCs 

relied on to meet the safety criteria 

defined in § 53.210 are protected against 

or designed to withstand the effects of 

external hazards under § 53.510. 

(B) The information necessary to 

demonstrate that the commercial 

nuclear plant complies with the 

earthquake engineering criteria in 

§ 53.480. 

(iii) A description of the defense-indepth measures required by § 53.250. 

(iv) A description of all plant 

operating states where there is the 

potential for the uncontrolled release of 

radioactive material to the environment, 

as required by § 53.450(b)(4). 

(v) A description of the events that 

challenge plant control and safety 

systems whose failure could lead to an 

undesirable end state and/or radioactive 

material release, as required by 

§ 53.450(b)(5). 

(vi) A description of the analytical 

codes used in modeling plant behavior 

in analyses of LBEs and how these 

codes are qualified for the range of 

conditions for which they were used, as 

required by § 53.450(d). 

(vii) If not described in addressing 

paragraph (5) of this section, the results 

of other analyses required by 

§ 53.450(g). 

(20) Special Treatments. A 

description of special treatments 

established as required by § 53.460. 

(21) Analytical Margins. A description 

of any alternative criteria adopted to 

demonstrate analytical margins 

supporting operational flexibilities, if 

applicable, as required by § 53.470. 

(22) Quality Assurance. A description 

of the QAP applied to the design of the 

SSCs of the commercial nuclear plant, 

as required by § 53.460(b). The 

description of the QAP for a commercial 

nuclear plant must include a discussion 

of how the applicable requirements of 

appendix B to part 50 of this chapter 

were satisfied. 

(23) Design Features and Controls to 

Address the Minimization of 

Contamination. The information 

required by § 20.1406 of this chapter. 

(24) Interface Requirements. (i) A 

description analysis, and evaluation of 

the interfaces between the standard 

design and the balance of the 

commercial nuclear plant that may 

impact the ability of the plant to 

demonstrate compliance with the 

functional design criteria or the safety 

criteria of subparts B and C of this part. 

(ii) Confirmation that interface 

requirements are verifiable through 

inspections, testing, or analysis. These 

requirements must be sufficiently 

detailed to allow for completion of the 

final safety analysis by license 

applicants that reference the certified 

design under this subpart. The method 

to be used for verification of interface 

requirements must be included as part 

of the proposed ITAAC required by 

§ 53.1241(a)(3). 


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(iii) A representative conceptual 

design for those portions of the plant for 

which the application does not seek 

certification to aid the NRC in its review 

of the FSAR and to permit assessment 

of the adequacy of the interface 

requirements under paragraph (a)(24)(i) 

of this section. 

(25) Technical Qualifications. A 

description of the technical 

qualifications of the applicant to engage 

in the proposed activities in accordance 

with the regulations in this chapter. 

(26) Technical Specifications. 

Proposed technical specifications 

prepared under § 53.710(a) for those 

areas addressed by the design 

certification. 

(27) Role of personnel. Information to 

address the following for the role of 

personnel in ensuring safe operations: 

(i) A description of how the human 

factors engineering design requirements 

of § 53.440(n)(1) are addressed; 

(ii) A description of how the human 

system interface design requirements of 

§ 53.440(n)(2) are addressed; 

(iii) A concept of operations that is of 

sufficient scope and detail to address 

the requirements of § 53.440(n)(3); 

(iv) A functional requirements 

analysis and function allocation that is 

of sufficient scope and detail to address 

the requirements of § 53.440(n)(4). 

(b) [Reserved] 

§ 53.1241 Contents of applications for 

standard design certifications; other 

application content. 

(a) In addition to the FSAR, the 

application must also include the 

following: 

(1) Environmental report. An 

environmental report as required by 

§ 51.55 of this chapter. 

(2) Availability Controls (if not 

included in the FSAR). A description of 

the controls on plant operations, 

including availability controls, to 

provide reasonable confidence that the 

configurations and special treatments 

for SR and NSRSS SSCs provide the 

capabilities and reliabilities required to 

demonstrate compliance with the safety 

criteria of § 53.220, or more restrictive 

alternative criteria adopted under 

§ 53.470. 

(3) Inspections, tests, analyses, and 

acceptance criteria. The proposed 

ITAAC that are necessary and sufficient 

to provide reasonable assurance that, if 

the inspections, tests, and analyses are 

performed and the acceptance criteria 

met, a facility that incorporates the 

design certification has been 

constructed and will be operated in 

conformity with the design certification, 

the provisions of the Act, and the 

Commission’s rules and regulations. 

(4) Safeguards information. A 

description of the program to protect 

Safeguards Information against 

unauthorized disclosure in accordance 

with the requirements in §§ 73.21 and 

73.22 of this chapter, as applicable. 

(b) If there are SSCs of the plant 

which required research and 

development to confirm the adequacy of 

their design, provide a report in the 

application which documents the 

resolution of any safety questions 

associated with such SSCs. 

(c) A description of how the 

performance of each design feature has 

been demonstrated capable of fulfilling 

functional design criteria considering 

interdependent effects through either 

analysis, appropriate test programs, 

prototype testing, operating experience, 

or a combination thereof, in accordance 

with § 53.440(a). 

§ 53.1242 Review of applications. 

(a) Standards for review of 

applications. Applications filed under 

this part will be reviewed for 

compliance with the standards set out 

in this part and 10 CFR parts 20, 51, and 

73. 

(b) Administrative review of 

applications; hearings. (1) A standard 

design certification is a rule that will be 

issued under the provisions of subpart 

H of 10 CFR part 2, as supplemented by 

the provisions of this section. The 

Commission must initiate the 

rulemaking after an application has 

been filed under § 53.1233 and must 

specify the procedures to be used for the 

rulemaking. The notice of proposed 

rulemaking published in the Federal 

Register must provide an opportunity 

for the submission of comments on the 

proposed design certification rule. If, at 

the time a proposed design certification 

rule is published in the Federal Register 

under this paragraph, the Commission 

decides that a legislative hearing should 

be held, the information required by 

§ 2.1502(c) of this chapter must be 

included in the Federal Register 

document for the proposed design 

certification. 

(2) Following the submission of 

comments on the proposed design 

certification rule, the Commission may, 

at its discretion, hold a legislative 

hearing under the procedures in subpart 

O of part 2 of this chapter. The 

Commission must publish a document 

in the Federal Register of its decision to 

hold a legislative hearing. The 

document must contain the information 

specified in § 2.1502(c) of this chapter 

and specify whether the Commission or 

a presiding officer will conduct the 

legislative hearing. 

(3) Notwithstanding anything in 

§ 2.390 of this chapter to the contrary, 

proprietary information will be 

protected in the same manner and to the 

same extent as proprietary information 

submitted in connection with 

applications for licenses, provided that 

the design certification will be 

published in chapter I of this title. 

(c) Reference to an issued operating 

license or combined license. In those 

cases where a design certification 

application is preceded by the issuance 

of an OL or custom COL for a 

commercial nuclear plant that is 

essentially the same as the standard 

design for which certification is being 

requested, the NRC review will follow 

the processes for referencing a standard 

design approval in § 53.1221, to the 

extent practicable. 

§ 53.1245 Referral to the Advisory 

Committee on Reactor Safeguards. 

The Commission must refer a copy of 

the application to the ACRS. The ACRS 

must report on those portions of the 

application which concern safety. 

§ 53.1248 Issuance of standard design 

certification. 

(a) After conducting a rulemaking 

proceeding under § 53.1242 on an 

application for a standard design 

certification and receiving the report to 

be submitted by the ACRS under 

§ 53.1245, the Commission may issue a 

standard design certification in the form 

of a rule for the design that is the subject 

of the application, if the Commission 

determines that— 

(1) The application demonstrates 

compliance with the applicable 

standards and requirements of the Act 

and the Commission’s regulations; 

(2) Notifications, if any, to other 

agencies or bodies have been duly 

made; 

(3) There is reasonable assurance that 

the standard design conforms with the 

provisions of the Act and the 

Commission’s regulations; 

(4) The applicant is technically 

qualified; 

(5) The proposed ITAAC are 

necessary and sufficient, within the 

scope of the standard design, to provide 

reasonable assurance that, if the 

inspections, tests, and analyses are 

performed and the acceptance criteria 

met, the facility has been constructed 

and will be operated in accordance with 

the design certification, the provisions 

of the Act, and the Commission’s 

regulations; 

(6) Issuance of the standard design 

certification will not be inimical to the 

common defense and security or to the 

health and safety of the public; 


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(7) The findings required by subpart 

A of part 51 of this chapter have been 

made; and 

(8) The applicant has implemented 

the QAP described or referenced in the 

Safety Analysis Report. 

(b) The design certification rule must 

specify the site parameters, design 

characteristics, and any additional 

requirements and restrictions of the 

design certification rule. 

(c) After the Commission has adopted 

a final design certification rule, the 

applicant must not permit any 

individual to have access to or any 

facility to possess restricted data or 

classified National Security Information 

until the individual and/or facility has 

been approved for access under the 

provisions of 10 CFR parts 25 and/or 95, 

as applicable. 

§ 53.1251 Duration of certification. 

(a) Except as provided in paragraph 

(b) of this section, a standard design 

certification issued under this subpart is 

valid for 15 years from the effective date 

of the rule. 

(b) A standard design certification 

continues to be valid beyond the date of 

expiration in any proceeding on an 

application for a COL or an OL under 

this part that references the standard 

design certification and is docketed 

either before the date of expiration of 

the certification, or, if a timely 

application for renewal of the 

certification has been filed, before the 

Commission has determined whether to 

renew the certification. A design 

certification also continues to be valid 

beyond the date of expiration in any 

hearing held under § 53.1452 before 

operation begins under a COL that 

references the design certification. 

(c) An applicant for a CP, OL, COL, 

or ML under this part may, at its own 

risk, reference in its application a design 

for which a design certification 

application has been docketed but not 

granted. 

§ 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 § 53.1251. 

(c) The Commission must refer a copy 

of the application for renewal to the 

ACRS. The ACRS must report on those 

portions of the application which 

concern safety and must apply the 

criteria set forth in § 53.1257. 

§ 53.1257 Criteria for renewal. 

(a) The Commission must issue a rule 

granting the renewal if the design, either 

as originally certified or as modified 

during the rulemaking on the renewal, 

complies with the Act and the 

Commission’s regulations applicable 

and in effect at the time the certification 

was issued. 

(b) The Commission may impose 

other requirements if it determines 

that— 

(1) They are necessary for adequate 

protection to public health and safety or 

common defense and security; 

(2) They are necessary for compliance 

with the Commission’s regulations and 

orders applicable and in effect at the 

time the design certification was issued; 

or 

(3) There is a substantial increase in 

overall protection of the public health 

and safety or the common defense and 

security to be derived from the new 

requirements, and the direct and 

indirect costs of implementing those 

requirements are justified in view of this 

increased protection. 

(c) In addition, the applicant for 

renewal may request an amendment to 

the design certification. The 

Commission must grant the amendment 

request if it determines that the 

amendment will comply with the Act 

and the Commission’s regulations in 

effect at the time of renewal. If the 

amendment request entails such an 

extensive change to the design 

certification that an essentially new 

standard design is being proposed, an 

application for a design certification 

must be filed in accordance with this 

subpart. 

(d) Denial of renewal does not bar the 

applicant, or another applicant, from 

filing a new application for certification 

of the design, which proposes design 

changes that correct the deficiencies 

cited in the denial of the renewal. 

§ 53.1260 Duration of renewal. 

Each renewal of certification for a 

standard design will be for not less than 

10, nor more than 15 years. 

§ 53.1263 Finality of standard design 

certifications. 

(a)(1) While a standard design 

certification rule is in effect under 

§§ 53.1251 or 53.1260, the Commission 

may not modify, rescind, or impose new 

requirements on the certification 

information, whether on its own 

motion, or in response to a petition from 

any person, unless the Commission 

determines in a rulemaking that the 

change— 

(i) Is necessary either to bring the 

certification information or the 

referencing plants into compliance with 

the Commission’s regulations applicable 

and in effect at the time the certification 

was issued; 

(ii) Is necessary to provide adequate 

protection of the public health and 

safety or the common defense and 

security; 

(iii) Reduces unnecessary regulatory 

burden and maintains protection to 

public health and safety and the 

common defense and security; 

(iv) Provides the detailed design 

information to be verified under those 

ITAAC that are directed at certification 

information (i.e., design acceptance 

criteria); 

(v) Is necessary to correct material 

errors in the certification information; 

(vi) Substantially increases overall 

safety, reliability, or security of facility 

design, construction, or operation, and 

the direct and indirect costs of 

implementation of the rule change are 

justified in view of this increased safety, 

reliability, or security; or 

(vii) Contributes to increased 

standardization of the certification 

information. 

(2)(i) In a rulemaking under 

§ 53.1263(a)(1), except for 

§ 53.1263(a)(1)(ii), the Commission will 

give consideration to whether the 

benefits justify the costs for plants that 

are already licensed or for which an 

application for a permit or license is 

under consideration. 

(ii) The rulemaking procedures for 

changes under § 53.1263(a)(1) must 

provide for notice and opportunity for 

public comment. 

(3) Any modification the NRC 

imposes on a design certification rule 


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under paragraph (a)(1) of this section 

will be applied to all plants referencing 

the certified design, except those to 

which the modification has been 

rendered technically irrelevant by 

action taken under paragraphs (a)(4) or 

(b) of this section. 

(4) The Commission may not impose 

new requirements by plant-specific 

order on any part of the design of a 

specific plant referencing the design 

certification rule if that part was 

approved in the design certification 

while a design certification rule is in 

effect under § 53.1248, unless— 

(i) A modification is necessary to 

secure compliance with the 

Commission’s regulations applicable 

and in effect at the time the certification 

was issued, or to assure adequate 

protection of the public health and 

safety or the common defense and 

security; and 

(ii) Special circumstances as defined 

in § 53.080 are present. In addition to 

the factors listed in § 53.080, the 

Commission must consider whether the 

special circumstances which § 53.080 

requires to be present outweigh any 

decrease in safety that may result from 

the reduction in standardization caused 

by the plant-specific order. 

(5) Except as provided in § 2.335 of 

this chapter, in making the findings 

required for issuance of a COL, CP, OL, 

or ML, or for any hearing under 

§ 53.1452, the Commission must treat as 

resolved those matters resolved in 

connection with the issuance or renewal 

of a design certification rule. 

(b) An applicant who references a 

design certification rule may request an 

exemption from one or more elements of 

the certification information. The 

Commission may grant such a request 

only if it determines that the exemption 

will comply with the requirements of 

§ 53.080. In addition to the factors listed 

in § 53.080, the Commission must 

consider whether the special 

circumstances that § 53.080 requires to 

be present outweigh any decrease in 

safety that may result from the 

reduction in standardization caused by 

the exemption. The granting of an 

exemption on request of an applicant is 

subject to litigation in the same manner 

as other issues in the OL or COL 

hearing. 

(c) The Commission will require, 

before granting a CP, COL, OL, or ML 

that references a design certification 

rule, that information normally 

contained in engineering documents, 

such as analyses, drawings, 

procurement specifications, or 

construction and installation 

specifications, be completed and 

available for audit if the more detailed 

information is necessary for the 

Commission to verify the information in 

the application and make its safety 

determination, including the 

determination that the application is 

consistent with the certification 

information. This information may be 

acquired by appropriate arrangements 

with the design certification applicant. 

§ 53.1270 Manufacturing licenses. 

Sections 53.1270 through 53.1295 set 

out the requirements and procedures 

applicable to Commission issuance of a 

license under this part authorizing 

manufacture of manufactured reactors to 

be installed at sites not identified in the 

ML application. 

§ 53.1273 Filing of applications. 

Any person, except one excluded by 

§ 53.1118, may file an application for an 

ML under this part with the Director, 

Office of Nuclear Reactor Regulation. 

§ 53.1276 Contents of applications for 

manufacturing licenses; general 

information. 

Each application for an ML must 

include the information contained in 

§ 53.1109(a) through (e), and (j). 

§ 53.1279 Contents of applications for 

manufacturing licenses; technical 

information. 

(a) Final Safety Analysis Report-siting 

and design. The application must 

include an FSAR containing the 

information set forth below, with a level 

of design information sufficient to 

enable the Commission to judge the 

applicant’s proposed means of ensuring 

that the manufacturing conforms to the 

design and to reach a final conclusion 

on all safety questions associated with 

the design, permit the preparation of 

construction and installation 

specifications by an applicant who 

seeks to use the manufactured reactor, 

and permit the preparation of 

acceptance and inspection requirements 

by the NRC. The application must 

include the following information: 

(1) Site parameters. The site 

parameters postulated for the design 

under this part, including the designbasis external hazard levels for the 

relevant external hazards, and an 

analysis and evaluation of the design in 

terms of those site parameters. 

(2) Design information. Except as 

specified in this paragraph, the design 

information equivalent to that required 

for a standard design certification as 

defined in § 53.1239(a)(2) through (27) 

for those portions of a commercial 

nuclear plant included in the 

manufactured reactor. 

(3) Quality assurance program. A 

description of the QAP applied to the 

design, and to be applied to the 

fabrication and testing of the SSCs of the 

manufactured reactor under 

§ 53.620(a)(6), including a discussion of 

how the applicable requirements of 

appendix B to part 50 of this chapter 

will be satisfied; 

(4) Conceptual designs. 

Representative conceptual designs for 

one or more commercial nuclear plants 

using the manufactured reactor; 

(5) Operating configurations. If 

multiple manufactured reactors may be 

installed at a commercial nuclear plant, 

a description of the possible operating 

configurations, including common 

systems, interface requirements, and 

system interactions. The final safety 

analysis must also account for 

differences among the possible 

configurations, including any 

restrictions that will be necessary 

during the construction and startup of a 

given manufactured reactor to ensure 

the safe operation of any commercial 

nuclear reactor already operating; 

(6) Interface requirements. (i) The 

interface requirements between the 

manufactured reactor and the remaining 

portions of the commercial nuclear 

plant or connections to other facilities 

outside of the commercial nuclear plant. 

(ii) Confirmation that interface 

requirements are verifiable through 

inspections, testing, or analysis. These 

requirements must be sufficiently 

detailed to allow for completion of the 

final safety analysis by license 

applicants that reference the 

manufactured reactor manufactured 

under this subpart. Applicants for a 

COL under this part will need to verify 

the interface requirements at the 

installation site. The method to be used 

for verification of interface requirements 

must be included as part of the 

proposed ITAAC required by 

§ 53.1282(a). 

(iii) Information to support 

development of radiation monitoring 

programs required under subpart F of 

this part by an applicant for a COL, 

including potential pathways for 

radionuclides produced within the 

manufactured reactor to enter 

interfacing systems. 

(b) Final Safety Analysis Report— 

manufacturing information. The FSAR 

must include the following information 

related to the manufacturing processes, 

organization, controls, and inspections: 

(1) A description, including 

references to generally accepted 

consensus codes and standards, of the 

processes that will be used to procure, 

fabricate, and assemble components that 

make up the manufactured reactor. The 

description should clearly define which 

activities are proposed to be within the 


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scope of the ML and those, such as the 

making of a component to be procured 

from a separate company for installation 

in the manufactured reactor, that are not 

considered to be within the scope of the 

ML; 

(2) A description of the organizational 

and management structure singularly 

responsible for direction of design and 

manufacture of the manufactured 

reactor. The information should include 

a description of the management plans, 

technical qualifications, and controls in 

place to demonstrate compliance with 

the requirements of § 53.620; 

(3) A description of the inspections 

and tests to be performed as part of the 

manufacturing process, including the 

inspection of procured components, 

inspection and testing of fabrication 

processes such as the molding, welding, 

or coating of components, and 

inspections and testing of the assembled 

manufactured reactor or portions of the 

manufactured reactor; 

(4) A description of the fitness-forduty program required by part 26 of this 

chapter and its implementation. 

(c) Deployment of the completed 

manufactured reactor. The application 

must include the following information 

related to the deployment of a 

manufactured reactor: 

(1) Procedures governing the 

preparation of the manufactured reactor 

or portions of the manufactured reactor 

for shipping to the site where it is to be 

operated; the conduct of shipping; and 

verifying the condition of the shipped 

items upon receipt at the site; 

(2) Details of the interaction of the 

design, manufacture, and installation of 

a manufactured reactor within the 

applicant’s organization and the manner 

by which the applicant will ensure close 

integration between the designer, 

contractors, and any facility in which 

the manufactured reactor is to be 

installed; 

(3) A description of the measures to 

be used for the control of interfaces, 

including the consideration of key site 

parameters, between the holder of the 

ML and the holder of the COL for the 

commercial nuclear plant at which the 

manufactured reactor is to be installed. 

(d) Special considerations for factory 

fueling. In addition to the above 

paragraphs, an application for an ML for 

a manufactured reactor that will be 

fueled at the factory under a 10 CFR part 

70 license must include the following 

information related to loading fuel and 

the required independent physical 

mechanisms to prevent criticality and to 

otherwise provide assurance that the 

fueled manufactured reactor can be 

successfully transported, installed, and 

operated at a site for which the 

Commission has issued a COL that 

authorizes construction and operation of 

a commercial nuclear plant using the 

manufactured reactor: 

(1) A description of the procedures 

used during the fueling of the 

manufactured reactor that ensure that 

the configuration of fuel within the 

fueled manufactured reactor is 

consistent with the design and analyses 

supporting operation of the 

manufactured reactor under the COL at 

the place of operation. The description 

may reference the applicable 10 CFR 

part 70 application and other sections of 

the Safety Analysis Report supporting 

the ML license application. 

(i) The application must describe the 

measures taken for in-factory 

inspections and non-nuclear testing 

performed to ensure that the 

configuration of fuel within the fueled 

manufactured reactor is consistent with 

the design and analyses supporting 

operation of the manufactured reactor 

under the COL at the place of operation. 

(ii) The application must describe the 

design features included in the 

manufactured reactor to prevent 

criticality, including at least two 

independent mechanisms each of which 

is sufficient to prevent criticality, the 

associated functional design criteria 

applied to those design features, and the 

physical and programmatic controls 

implemented during manufacturing, 

storage, and transport that are credited 

to assure the features function as 

designed when subject to potential 

hazards and human errors. The 

descriptions must include how those 

measures will be controlled during 

installation under the ML and removal 

under the COL at the place of operation. 

(2) A description of the procedures 

governing the transfer of responsibilities 

for the fueled manufactured reactor 

from the holder of the ML to the holder 

of the COL for the installation site. 

(3) If available at the time of filing the 

ML application or, if not available at the 

time of filing the ML application, 

submitted as an amendment to the ML 

or ML application at the time of filing 

the Part 70 application, a description of 

the programs needed to demonstrate 

compliance with the requirements of 

§ 53.620(d) and 10 CFR parts 70, 71, and 

73 for the receipt, storage, and loading 

of SNM into a manufactured reactor and 

the transport of the fueled manufactured 

reactor to a site for which the 

Commission has issued a COL that 

authorizes construction and operation of 

a commercial nuclear plant using the 

manufactured reactor, including the 

following. 

(i) A physical security program in 

accordance with § 53.620(d)(2)(i). 

(ii) A cybersecurity program in 

accordance with § 53.620(d)(2)(i). 

§ 53.1282 Contents of applications for 

manufacturing licenses; other application 

content. 

(a) Inspections, tests, analyses, and 

acceptance criteria. (1) The application 

must contain proposed inspections, 

tests, and analyses that the COL holder 

must perform, and the acceptance 

criteria that are necessary and sufficient 

to provide reasonable assurance that, if 

the inspections, tests, and analyses are 

performed and the acceptance criteria 

met: 

(i) The reactor has been manufactured 

in conformity with the ML, the 

provisions of the Act, and the 

Commission’s rules and regulations; and 

(ii) The manufactured reactor will be 

operated in conformity with the 

approved design and any license 

authorizing operation of the 

manufactured reactor. 

(2) If the application references a 

standard design certification, the ITAAC 

contained in the certified design must 

apply to those portions of the facility 

design that are covered by the design 

certification. 

(3) If the application references a 

standard design certification, the 

application may include a notification 

that a required inspection, test, or 

analysis in the design certification 

ITAAC has been successfully completed 

and that the corresponding acceptance 

criterion has been met. The Federal 

Register notification required by 

§ 53.1285 must indicate that the 

application includes this notification. 

(b) Environmental report. (1) The 

application must contain an 

environmental report as required by 

§ 51.54 of this chapter. 

(2) If the ML application references a 

standard design certification, the 

environmental report need not contain a 

discussion of severe accident mitigation 

design alternatives for the manufactured 

reactor as used in a commercial nuclear 

plant. 

(c) Safeguards information. The 

application must contain a description 

of the program to protect safeguards 

information against unauthorized 

disclosure in accordance with the 

requirements in §§ 73.21 and 73.22 of 

this chapter, as applicable. 

(d) Performance demonstration. A 

description of how the performance of 

each design feature has been 

demonstrated capable of fulfilling 

functional design criteria considering 

interdependent effects through either 

analysis, appropriate test programs, 

prototype testing, operating experience, 


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or a combination thereof, in accordance 

with § 53.440(a). 

§ 53.1285 Review of applications. 

(a) Standards for review of 

applications. Applications for MLs 

under this part will be reviewed 

according to the applicable standards 

set out in this subpart as well as 

applicable standards in this part and 10 

CFR parts 20, 25, 26, 51, 70, 71, 73, and 

75. 

(b) Administrative review of 

applications, hearings. A proceeding on 

an ML is subject to all applicable 

procedural requirements contained in 

10 CFR part 2, including the 

requirements for docketing in 

§ 2.101(a)(1) through (4) of this chapter, 

and the requirements for issuance of a 

notice of proposed action in § 2.105 of 

this chapter, provided, however, that the 

designated sections may not be 

construed to require that the 

environmental report or draft or final 

environmental impact statement include 

an assessment of the benefits of 

constructing and/or operating the 

manufactured reactor or an evaluation 

of alternative energy sources. All 

hearings on MLs are governed by the 

hearing procedures contained in 10 CFR 

part 2, subparts C, E, G, L, and N. 

§ 53.1286 Referral to the Advisory 

Committee on Reactor Safeguards. 

The Commission must refer a copy of 

the application to the ACRS. The ACRS 

must report on those portions of the 

application which concern safety. 

§ 53.1287 Issuance of manufacturing 

licenses. 

(a) After completing any hearing 

under § 53.1285(b), and receiving the 

report submitted by the ACRS, the 

Commission may issue an ML if the 

Commission finds that— 

(1) Applicable standards and 

requirements of the Act and the 

Commission’s regulations have been 

met; 

(2) There is reasonable assurance that 

the manufactured reactor will be 

manufactured, and can be transported, 

incorporated into a commercial nuclear 

plant, and operated in conformity with 

the ML, the provision of the Act, and 

the Commission’s regulations; 

(3) The proposed manufactured 

reactor can be incorporated into a 

commercial nuclear plant and operated 

at sites having characteristics that fall 

within the site parameters postulated for 

the design of the manufactured reactors 

without undue risk to the health and 

safety of the public; 

(4) The applicant is technically 

qualified to design and manufacture the 

proposed manufactured reactor; 

(5) The proposed ITAAC are 

necessary and sufficient, within the 

scope of the ML, to provide reasonable 

assurance that the manufactured reactor 

has been manufactured and will be 

operated in conformity with the license, 

the provisions of the Act, and the 

Commission’s regulations; 

(6) The issuance of a license to the 

applicant will not be inimical to the 

common defense and security or to the 

health and safety of the public; and 

(7) The findings required by subpart 

A of 10 CFR part 51 have been made. 

(b) Each ML issued under this subpart 

must specify— 

(1) Terms and conditions as the 

Commission deems necessary and 

appropriate; 

(2) Technical specifications for 

operation of the manufactured reactor, 

as the Commission deems necessary and 

appropriate; 

(3) Site parameters and design 

characteristics for the manufactured 

reactor; 

(4) The interface requirements to be 

met by the site-specific elements of the 

facility, such as the energy conversions 

systems and ultimate heat sink, not 

within the scope of the manufactured 

reactor; and 

(5) The entity with design authority 

for the manufactured reactor covered by 

the license. 

§ 53.1288 Finality of manufacturing 

licenses. 

(a)(1) Notwithstanding any provision 

in § 53.1590, during the term of an ML 

issued under this part the Commission 

may not modify, rescind, or impose new 

requirements on the design of the 

manufactured reactor, or the 

requirements for the manufacture of the 

manufactured reactor, unless the 

Commission determines that a 

modification is necessary to bring the 

design of the reactor or its manufacture 

into compliance with the Commission’s 

requirements applicable and in effect at 

the time the ML was issued, or to 

provide reasonable assurance of 

adequate protection to public health and 

safety or common defense and security. 

(2) Any modification to the design of 

a manufactured reactor that is imposed 

by the Commission under paragraph 

(a)(1) of this section will be applied to 

all manufactured reactors manufactured 

under the license, including those that 

have already been transported and sited, 

except those manufactured reactors to 

which the modification has been 

rendered technically irrelevant by 

action taken under § 53.1530 or 

paragraph (b) of this section. 

(3) In making the findings required 

under this part for issuance of a COL, 

in any hearing under § 53.1452, or in 

any enforcement hearing other than one 

initiated by the Commission under 

paragraph (a)(1) of this section, for 

which a manufactured reactor 

manufactured under this subpart is 

referenced or used, the Commission 

must treat as resolved those matters 

resolved in the proceeding on the 

application for issuance or renewal of 

the ML, including the adequacy of 

design of the manufactured reactor, the 

costs and benefits of severe accident 

mitigation design alternatives, and the 

bases for not incorporating severe 

accident mitigation design alternatives 

into the design of the manufactured 

reactor to be manufactured. 

(b) An applicant who references or 

uses a manufactured reactor 

manufactured under an ML under this 

part may include in the application a 

request for a departure from the design 

characteristics, site parameters, terms 

and conditions, or approved design of 

the manufactured reactor. The 

Commission may grant a request only if 

it determines that the departure will 

comply with the requirements of 

§ 53.080, and that the special 

circumstances outweigh any decrease in 

safety that may result from the 

reduction in standardization caused by 

the departure. The granting of a 

departure on request of an applicant is 

subject to litigation in the same manner 

as other issues in the COL hearing. 

§ 53.1291 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. 

§ 53.1293 Transfer of manufacturing 

licenses. 

An ML may be transferred under 

§ 53.1570. 

§ 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 

§ 53.1100. 


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(3) An ML issued under this part, 

either original or renewed, for which a 

timely application for renewal has been 

filed, remains in effect until the 

Commission has made a final 

determination on the renewal 

application, provided, however, that the 

holder of an ML may not begin 

manufacture of a manufactured reactor 

less than 6 months before the expiration 

of the license. 

(4) Any person whose interest may be 

affected by renewal of the license may 

request a hearing on the application for 

renewal. The request for a hearing must 

comply with § 2.309 of this chapter. If 

a hearing is granted, notice of the 

hearing will be published in accordance 

with § 2.104 of this chapter. 

(5) The Commission must refer a copy 

of the application for renewal to the 

ACRS. The ACRS must report on those 

portions of the application which 

concern safety. 

(b) The Commission may grant the 

renewal if the Commission 

determines— 

(1) The ML complies with the Act and 

the Commission’s regulations and 

orders applicable and in effect at the 

time the ML was originally issued; and 

(2) Any new requirements the 

Commission may wish to impose are— 

(i) Necessary for adequate protection 

to public health and safety or common 

defense and security; 

(ii) Necessary for compliance with the 

Commission’s regulations and orders 

applicable and in effect at the time the 

ML was originally issued; or 

(iii) A substantial increase in overall 

protection of the public health and 

safety or the common defense and 

security to be derived from the new 

requirements, and the direct and 

indirect costs of implementation of 

those requirements are justified in view 

of this increased protection. 

(c) A renewed ML may be issued for 

a term of not less than 5, nor more than 

15 years, plus any remaining years on 

the ML then in effect before renewal. 

The renewed license must be subject to 

the requirements of § 53.1288. 

§ 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 

§§ 53.1360 through 53.1405. 

§ 53.1306 Contents of applications for 

construction permits; general information. 

An application for a CP must include 

the information required by § 53.1109 

and the following information: 

(a) Information sufficient to 

demonstrate to the Commission the 

financial qualification of the applicant 

to carry out, under the regulations in 

this chapter, the activities for which the 

permit is sought. As applicable, the 

following should be provided: 

(1) The information that demonstrates 

that the applicant possesses or has 

reasonable assurance of obtaining the 

funds necessary to cover estimated 

construction costs and related fuel cycle 

costs, including estimates of the total 

construction costs and related fuel cycle 

costs of the facility and must indicate 

the source(s) of funds to cover these 

costs. 

(2) Each application for a CP 

submitted by a newly formed entity 

organized for the primary purpose of 

constructing and operating a facility 

must also include information showing: 

(i) The legal and financial 

relationships the entity has or proposes 

to have with its stockholders or owners; 

(ii) The stockholders’ or owners’ 

financial ability to meet any contractual 

obligation to the entity that they have 

incurred or proposed to incur; and 

(iii) Any other information considered 

necessary by the Commission to enable 

it to determine the applicant’s financial 

qualification; and 

(3) The Commission may request an 

established entity or newly-formed 

entity to submit additional or more 

detailed information respecting its 

financial arrangements and status of 

funds if the Commission considers this 

information appropriate. This may 

include information regarding an 

applicant’s ability to continue the 

conduct of the activities authorized by 

the CP and to decommission the facility. 

(b) If the applicant proposes to 

construct or alter a facility, the 

application must state the earliest and 

latest dates for completion of the 

construction or alteration. 

§ 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 

§ 53.1146(a)(1)(iv) through (x). 

(2) Design information. Except as 

specified in this paragraph, an 

application for a CP for a commercial 

nuclear plant must include the design 

information equivalent to that required 

for a standard design certification as 

defined in § 53.1239(a)(2) through (27). 

(i) Quality assurance program. A 

description of the QAP to be applied to 

the design, fabrication, construction, 

and testing of the SSCs of the facility 

under § 53.610(a)(6), including a 

discussion of how the requirements of 

appendix B to part 50 of this chapter 

will be satisfied. 

(ii) Preliminary design information. 

The information provided in the 

application may include some aspects of 

the design that are not fully developed, 

and the information is therefore 

preliminary. The completed design, 

including any changes during 

construction, must be described in the 

FSAR required in § 53.1369 that 

supports an application for an OL. 

(iii) Planned research or testing. 

Descriptions of how design features and 

related functional design criteria will 

fulfill the safety criteria in subpart B, or 

more restrictive alternative criteria 

adopted under § 53.470, and how that 

has been or will be demonstrated 

through either analysis, appropriate test 

programs, experience, or a combination 

thereof. Where any design feature has 

not been fully developed or 

demonstrated to fulfill the functional 

design criteria at the time of an 

application for a CP, the applicant must 

provide a plan for future analysis, 

research and development, test 

programs, gathering of experience, or a 

combination thereof to provide 

reasonable confidence that the required 

demonstration will be available for an 

application for an OL. 

(iv) Programmatic controls. 

Descriptions of the programmatic 

controls may include those to be 

provided in the FSAR or other licensing 

basis documents because they are 

necessary to achieve and maintain the 

reliability and capability of SSCs relied 

upon to demonstrate compliance with 

the established safety criteria and 

functional design criteria required in 

subpart B, and to maintain consistency 

with analyses required by § 53.450. 

(3) Technical qualifications. A 

description of the technical 

qualifications of the applicant to engage 

in the proposed activities under the 

regulations in this chapter. 


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(4) Emergency preparedness. A 

description of the applicant’s 

preliminary plans for coping with 

emergencies based on: 

(i) Except as provided in paragraph 

(a)(4)(ii) of this section, the 

requirements in appendix E to part 50. 

(ii) For a commercial nuclear plant 

consisting of either small modular 

reactors or non-light-water reactors, the 

requirements in either § 50.160 or 

appendix E to part 50. 

(5) Physical security. A report that 

provides a preliminary description of 

how the site characteristics support the 

development of adequate security plans 

and measures consistent with the 

requirements in § 53.540. 

(6) Fitness-for-duty program. A 

description of the fitness-for-duty (FFD) 

program required by 10 CFR part 26 and 

its implementation. 

(b) A description of the program to 

protect Safeguards Information against 

unauthorized disclosure in accordance 

with the requirements in §§ 73.21 and 

73.22 of this chapter, as applicable. 

§ 53.1312 Contents of applications for 

construction permits; other application 

content. 

(a) In addition to the PSAR, the 

application must include the following: 

(1) An environmental report either 

under § 51.50(a) of this chapter if an 

LWA under § 53.1130 is not requested 

in conjunction with the CP application, 

or under §§ 51.49 and 51.50(a) of this 

chapter if an LWA is requested in 

conjunction with the CP application; or 

(2) If the applicant wishes to request 

that an LWA under § 53.1130 be issued 

before issuance of the CP, the 

information otherwise required by 

§ 53.1130, in accordance with either 

§ 2.101(a)(1) through (a)(5), or 

§ 2.101(a)(9) of this chapter. 

(b) If the CP application references an 

early site permit, standard design 

approval, or standard design 

certification issued under this part, then 

the following requirements apply: 

(1) The PSAR need not contain 

information or analyses submitted to the 

Commission in connection with the 

referenced NRC approval, permit, or 

certification, provided, however, that 

the PSAR incorporates the material by 

reference and confirms that the site and 

design of the facility falls within 

parameter values postulated in the 

referenced NRC approval, permit, or 

certification. 

(2) The PSAR must provide a means 

to demonstrate that all terms and 

conditions that have been included in 

the referenced NRC approval, permit, or 

certification will be satisfied by the date 

of issuance of the OL, as appropriate. If 

the PSAR does not demonstrate that 

each site characteristic falls within the 

corresponding postulated site parameter 

and each design characteristic of the 

facility falls within the corresponding 

postulated design parameter, the 

application must justify a departure, 

variance, or exemption from the 

referenced NRC approval, license, or 

certification in regard to that particular 

site or design characteristic in 

compliance with the requirements of 

this part. 

(3) If a referenced early site permit 

approves complete and integrated 

emergency plans, or major features of 

emergency plans, then the PSAR must 

include any new or additional 

information that updates and corrects 

the information that was provided 

under § 53.1146(b)(2) and discuss 

whether the new or additional 

information materially changes the 

bases for compliance with the 

applicable requirements. 

§ 53.1315 Review of applications. 

(a) Standards for review of 

applications. Applications filed under 

this part will be reviewed according to 

the standards set out in this part and 10 

CFR parts 20, 51, 73, and 140. 

(b) Administrative review of 

applications; hearings. A proceeding on 

a CP application is subject to all 

applicable procedural requirements 

contained in 10 CFR part 2, including 

the requirements for docketing (§ 2.101 

of this chapter) and issuance of a notice 

of hearing (§ 2.104 of this chapter). All 

hearings on CP applications are 

governed by the procedures contained 

in 10 CFR part 2. 

§ 53.1318 Finality of referenced NRC 

approvals, permits, and certifications. 

If the application for a CP under this 

part references an early site permit, 

standard design approval, or standard 

design certification, the scope and 

nature of matters resolved for the 

application are governed by the relevant 

provisions addressing finality, including 

§§ 53.1188, 53.1221, and 53.1263. 

§ 53.1324 Referral to the Advisory 

Committee on Reactor Safeguards. 

The Commission must refer a copy of 

the application to the ACRS. The ACRS 

must report on those portions of the 

application that concern safety and 

must apply the standards referenced in 

§ 53.1315, in accordance with the 

finality provisions in § 53.1318. 

§ 53.1327 Authorization to conduct limited 

work authorization activities. 

(a) If the application does not 

reference an early site permit which 

authorizes the holder to perform the 

activities under § 53.1130, the applicant 

may not perform those activities 

without obtaining the separate 

authorization required by § 53.1130. 

Authorization may be granted only after 

the presiding officer in the proceeding 

on the application has made the 

findings and determination required by 

§ 53.1130(b)(1)(ii) and (iv), and the 

Director, Office of Nuclear Reactor 

Regulation makes the determination 

required by § 53.1130(b)(1)(iii). 

(b) If, after an applicant has performed 

the activities permitted by paragraph (a) 

of this section, the application for the 

CP is withdrawn or denied, then the 

applicant must implement an approved 

site redress plan. 

§ 53.1330 Exemptions, departures, and 

variances. 

(a) Applicants for a CP under this 

part, or any amendment to a CP, may 

include in the application a request for 

an exemption from one or more of the 

Commission’s regulations. The 

Commission may grant a request if it 

determines that the exemption complies 

with § 53.080. 

(b) An applicant for a CP who has 

filed an application referencing an NRC 

approval, permit, or certification issued 

under this part may include in the 

application a request for exemptions, 

departures, or variances related to the 

subject referenced NRC approval, 

permit, or certification. In determining 

whether to grant the departure, 

variance, or exemption, the Commission 

must apply the same technically 

relevant criteria as were applicable to 

the application for the original or 

renewed approval, license, or 

certification. 

§ 53.1333 Issuance of construction 

permits. 

(a) After conducting a hearing in 

accordance with § 53.1315 and receiving 

the report submitted by the ACRS, the 

Commission may issue a CP only if the 

Commission finds that— 

(1) The applicant has described the 

proposed design of the facility and has 

identified the major features or 

components incorporated therein for the 

protection of the health and safety of the 

public; 

(2) Such further technical or design 

information as may be required to 

complete the safety analysis, and which 

can reasonably be left for later 

consideration, will be supplied in the 

FSAR; 

(3) Safety features or components, if 

any, that require research and 

development have been described by 

the applicant and the applicant has 

identified, and there will be conducted, 


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a research and development program 

reasonably designed to resolve any 

safety questions associated with such 

features or components; and 

(4) On the basis of the foregoing, there 

is reasonable assurance of the 

following— 

(i) Such safety questions will be 

satisfactorily resolved at or before the 

latest date stated in the application for 

completion of construction of the 

proposed facility; and 

(ii) Taking into consideration the site 

criteria contained subpart D to this part, 

the proposed facility can be constructed 

and operated at the proposed location 

without undue risk to the health and 

safety of the public. 

(b) A CP must contain the terms and 

conditions for the permit, as the 

Commission deems necessary and 

appropriate. The Commission may, in 

its discretion, incorporate in any CP 

provisions requiring the applicant to 

furnish periodic reports of the progress 

and results of research and development 

programs designed to resolve safety 

questions. 

§ 53.1336 Finality of construction permits. 

Notwithstanding any provision in 

§ 53.1590, a CP constitutes an 

authorization to proceed with 

construction but does not constitute 

Commission approval of the safety of 

any design feature or specification 

unless the applicant specifically 

requests such approval and such 

approval is incorporated in the permit. 

The applicant, at its option, may request 

such approvals in the CP or by 

amendment to the CP. If approved by 

the NRC and included in the permit, the 

NRC will consider modifications to the 

approved design features or 

specifications in accordance with 

§ 53.1590. 

§ 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. 

§ 53.1345 Transfer of construction 

permits. 

A CP may be transferred under 

§ 53.1570. 

§ 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. 

§ 53.1360 Operating licenses. 

Sections 53.1360 through 53.1405 set 

out the requirements and procedures 

applicable to Commission issuance of 

an OL for a nuclear power facility. 

§ 53.1366 Contents of applications for 

operating licenses; general information. 

An application for an OL must 

include the information required by 

§ 53.1109 and the following 

information: 

(a) Except for an electric utility 

applicant, information sufficient to 

demonstrate to the Commission the 

financial qualification of the applicant 

to carry out, in accordance with the 

regulations in this chapter, the activities 

for which the license is sought. As 

applicable, the following should be 

provided: 

(1) The applicant must submit 

information that demonstrates the 

applicant possesses or has reasonable 

assurance of obtaining the funds 

necessary to cover estimated operation 

costs for the period of the license. The 

applicant must submit estimates for 

total annual operating costs for each of 

the first 5 years of operation of the 

facility. The applicant must also 

indicate the source(s) of funds to cover 

these costs. 

(2) Each application for an OL 

submitted by a newly-formed entity 

organized for the primary purpose of 

operating the facility must also include 

information showing— 

(i) The legal and financial 

relationships the entity has or proposes 

to have with its stockholders or owners; 

(ii) The stockholders’ or owners’ 

financial ability to meet any contractual 

obligation to the entity which they have 

incurred or proposed to incur; and 

(iii) Any other information considered 

necessary by the Commission to enable 

it to determine the applicant’s financial 

qualification. 

(3) The Commission may request an 

established entity or newly formed 

entity to submit additional or more 

detailed information respecting its 

financial arrangements and status of 

funds if the Commission considers this 

information appropriate. This may 

include information regarding a 

licensee’s ability to continue the 

conduct of the activities authorized by 

the license and to decommission the 

facility. 

(b) The application must include 

information in the form of a report, as 

described in subpart G, indicating how 

reasonable assurance will be provided 

that funds will be available to 

decommission the facility, including a 

copy of the financial instrument 

obtained to satisfy the requirements of 

§ 53.1040. 

§ 53.1369 Contents of applications for 

operating licenses; technical information. 

Final Safety Analysis Report. The 

application must contain an FSAR that 

describes the facility and the limits on 

its operation and presents a safety 

analysis of the SSCs of the facility as a 

whole. The FSAR must include the 

following information, at a level of 

detail sufficient to enable the 

Commission to reach a final conclusion 

on all safety matters that must be 

resolved by the Commission before 

issuance of an OL. The FSAR must 

include the following information: 

(a) Site information. An application 

for an OL for a commercial nuclear 

reactor must include the site 

information equivalent to that required 

for an early site permit in 

§ 53.1146(a)(1)(iv) through (x), including 

all current information, such as the 

results of environmental and 

meteorological monitoring programs, 

which has been developed since 

issuance of the CP, relating to site 

evaluation factors identified in this part. 

(b) Design information. Except as 

specified in this paragraph, an FSAR for 

an OL for a commercial nuclear plant 

must include the final design 

information equivalent to that required 

for a standard design certification as 

defined in § 53.1239(a)(2) through (7), 

(a)(9), and (a)(11) through (a)(27). 

(1) The completed design, including 

any changes during construction, must 

be described. 

(2) Where any design feature had not 

been fully developed or demonstrated at 

the time of application for the CP, the 

applicant must provide the analysis, 

research and development, test 

programs, gathering of experience, or a 

combination thereof to provide the 

required demonstration to fulfill the 

functional design criteria. 

(c) Technical qualifications. A 

description of the technical 

qualifications of the applicant to engage 

in the proposed activities in accordance 

with the regulations in this chapter. 


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(d) Integrity assessment program. A 

description of an Integrity Assessment 

Program that addresses the elements 

described in § 53.870. 

(e) Safeguards information. A 

description of the program to protect 

Safeguards Information against 

unauthorized disclosure in accordance 

with the requirements in §§ 73.21 and 

73.22 of this chapter, as applicable. 

(f) Emergency response facility or 

facilities. Description of location and 

capabilities to be established for 

command and control, support, and 

coordination of onsite and offsite, as 

applicable, functions during reactor 

accident conditions. 

(g) Role of personnel. (1) A 

description of the completed 

assessments related to the role of 

personnel in ensuring safe operations 

considering the analyses required by 

§ 53.730. These assessments must 

include the following: 

(i) Human factors engineering design 

requirements of § 53.730(a); 

(ii) Human system interface design 

requirements of § 53.730(b); 

(iii) Concept of operations of 

§ 53.730(c); 

(iv) Functional requirements analysis 

and function allocation of § 53.730(d); 

(2) A description of the program to be 

used for evaluating and applying 

operating experience as required by 

§ 53.730(e); 

(3) A staffing plan and supporting 

analyses as required by § 53.730(f). 

(h) Training, examination, and 

proficiency programs. (1) A description 

of the training, examination, and 

proficiency programs required by 

§ 53.730(g); 

(2) A description of the training 

programs required by § 53.830. 

(i) Emergency plan. Emergency plans 

complying with the requirements of 

§ 53.855. 

(1) Include all emergency plan 

certifications, as applicable, that have 

been obtained from the State, local, and 

participating Tribal governmental 

agencies with emergency planning 

responsibilities that are wholly or 

partially within the EPZ plume 

exposure pathway. These certifications 

must state that— 

(i) The proposed emergency plans are 

practicable; 

(ii) These agencies are committed to 

participating in any further 

development of the plans, including any 

required field demonstrations; and 

(iii) These agencies are committed to 

executing their responsibilities under 

the plans in the event of an emergency. 

(2) If certifications cannot be obtained 

after sustained, good faith efforts by the 

applicant, then the application must 

contain information, including a utility 

plan, sufficient to show that the 

proposed plans provide reasonable 

assurance that adequate protective 

measures can and will be taken in the 

event of a radiological emergency at the 

site. 

(3) If complete and integrated 

emergency plans were approved as part 

of an early site permit, or submitted, 

reviewed, and approved as part of the 

CP application, new certifications that 

demonstrate compliance with the 

requirements of paragraph (i)(1) of this 

section are not required. 

(j) Organization. A description of the 

applicant’s organizational structure, 

allocations of responsibilities and 

authorities, and personnel qualifications 

requirements for operation. 

(k) Maintenance program. A 

description of a maintenance program 

under § 53.715. 

(l) Quality assurance. A description of 

the QAP that demonstrates compliance 

with the requirements under § 53.865. 

(m) Radiation protection program. A 

radiation protection program 

description under § 53.850. 

(n) Security program. A physical 

security plan that describes how the 

applicant will comply with § 53.860 

(and 10 CFR part 11, if applicable, 

including the identification and 

description of jobs as required by 

§ 11.11(a) of this chapter, at the 

proposed facility). The plan must list 

tests, inspections, audits, and other 

means to be used to demonstrate 

compliance with the requirements of 10 

CFR parts 11 and 73, if applicable. 

(o) Safeguards contingency plan. A 

safeguards contingency plan in 

accordance with the criteria set forth in 

appendix C to 10 CFR part 73. The 

safeguards contingency plan must 

include plans for dealing with threats, 

thefts, and radiological sabotage, as 

defined in 10 CFR part 73, relating to 

the SNM and nuclear facilities licensed 

under this chapter and in the 

applicant’s possession and control. Each 

application for this type of license must 

include the information contained in 

the applicant’s safeguards contingency 

plan. (Implementing procedures 

required for this plan need not be 

submitted for approval.) 1

(p) Security training and 

qualification. A training and 

qualification plan that describes how 

the applicant will demonstrate 

compliance with the criteria set forth in 

§ 73.100 of this chapter or appendix B 

to 10 CFR part 73. 

(q) Cybersecurity plan. A 

cybersecurity plan in accordance with 

the criteria set forth in § 73.54 or 

§ 73.110 of this chapter. 

(r) Security, safeguards and 

cybersecurity plan implementation. A 

description of the implementation of the 

physical security plan, safeguards 

contingency plan, training and 

qualification plan, and cybersecurity 

plan. Each applicant who prepares a 

physical security plan, a safeguards 

contingency plan, a training and 

qualification plan, or a cybersecurity 

plan must protect the plans and other 

related Safeguards Information against 

unauthorized disclosure in accordance 

with the requirements of §§ 73.21 and 

73.22 of this chapter. 

(s) Fire protection program. A 

description of the fire protection 

program under § 53.875. 

(t) Inservice inspection/inservice 

testing program. A description of the 

inservice inspection and inservice 

testing programs under § 53.880. 

(u) [Reserved] 

(v) [Reserved] 

(w) General employee training. A 

description of the training program 

required to demonstrate compliance 

with § 53.830 and its implementation. 

(x) Fitness-for-duty program. A 

description of the FFD program required 

by 10 CFR part 26 and its 

implementation. 

(y) Other programs. A description and 

evaluation of the results of the 

applicant’s programs, including 

research and development, if any, to 

demonstrate that any safety questions 

identified at the CP stage have been 

resolved. 

(z) Safety design feature performance. 

A description of how the performance of 

each safety design feature has been 

demonstrated capable of fulfilling 

functional design criteria considering 

interdependent effects through either 

analysis, appropriate test programs, 

prototype testing, operating experience, 

or a combination thereof, in accordance 

with § 53.440(a). 

(aa) Technical specifications. 

Proposed technical specifications 

prepared in accordance with the 

requirements of § 53.710(a). 

1A physical security plan that contains all 

the information required in both § 73.55 or 

§ 73.100 of this chapter and appendix C to 10 

CFR part 73 satisfies the requirement for a 

contingency plan. 

§ 53.1372 Contents of applications for 

operating licenses; other application 

content. 

In addition to the FSAR, the 

application must also include the 

following: 

(a) Environmental report. An 

environmental report in accordance 

with § 51.53(b) of this chapter. 

(b) Availability controls (if not 

included in the FSAR). A description of 


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the controls on plant operations, 

including availability controls, to 

provide reasonable confidence of safe 

operation and that the configurations 

and special treatments for SR and 

NSRSS SSCs provide the capabilities 

and reliabilities required to satisfy the 

safety criteria of § 53.220, or more 

restrictive alternative criteria adopted 

under § 53.470, if not addressed by 

Technical Specifications under 

§ 53.1369(aa). 

§ 53.1375 Review of applications. 

(a) Standards for review of 

applications. Applications filed under 

this part will be reviewed according to 

the standards set out in 10 CFR parts 20, 

26, 51, 53, 73, and 140. 

(b) Administrative review of 

applications; hearings. A proceeding on 

an OL is subject to all applicable 

procedural requirements contained in 

10 CFR part 2, including the 

requirements for docketing (§ 2.101 of 

this chapter) and issuance of a notice of 

hearing (§ 2.104 of this chapter). All 

hearings on OLs are governed by the 

procedures contained in 10 CFR part 2. 

§ 53.1381 Referral to the Advisory 

Committee on Reactor Safeguards. 

The Commission must refer a copy of 

the application to the ACRS. The ACRS 

must report on those portions of the 

application that concern safety and 

must apply the standards referenced in 

§ 53.1375. 

§ 53.1384 Exemptions, departures, and 

variances. 

(a) Applicants for an OL under this 

part, or any amendment to an OL, may 

include in the application a request for 

an exemption from one or more of the 

Commission’s regulations. The 

Commission may grant an exemption 

request if it determines that the 

exemption complies with § 53.080. 

(b) An applicant for an OL who has 

filed an application referencing an NRC 

approval, permit, license, or 

certification issued under this part may 

include in the application a request for 

departures, variances, or exemptions 

related to the subject referenced NRC 

approval, permit, license, or 

certification. In determining whether to 

grant the departure, variance, or 

exemption, the Commission must apply 

the same technically relevant criteria as 

were applicable to the application for 

the original or renewed approval, 

license, or certification. 

§ 53.1387 Issuance of operating licenses. 

Upon completion of the construction 

or alteration of a facility, in compliance 

with the terms and conditions of the 

construction permit and subject to any 

necessary testing of the facility for 

health or safety purposes, the 

Commission will, in the absence of good 

cause shown to the contrary, issue an 

OL or an appropriate amendment of the 

license, as the case may be. 

(a)(1) After receiving the report 

submitted by the ACRS, the 

Commission may issue an OL if the 

Commission finds that— 

(i) Construction of the facility has 

been substantially completed in 

conformity with the CP and the 

application as amended, the provisions 

of the Act, and the rules and regulations 

of the Commission; 

(ii) Any required notifications to other 

agencies or bodies have been duly 

made; 

(iii) The facility will operate in 

conformity with the application as 

amended, the provisions of the Act, and 

the rules and regulations of the 

Commission; 

(iv) There is reasonable assurance 

that— 

(A) the activities authorized by the OL 

can be conducted without endangering 

the health and safety of the public; and 

(B) such activities will be conducted 

in compliance with the regulations in 

this chapter. 

(v) The applicant is technically and 

financially qualified to engage in the 

activities authorized, however, no 

finding of financial qualification is 

necessary for an electric utility 

applicant for an OL; 

(vi) Issuance of the license will not be 

inimical to the common defense and 

security or to the health and safety of 

the public; 

(vii) The applicable provisions of 10 

CFR part 140 have been satisfied; and 

(viii) The findings required by subpart 

A of 10 CFR part 51 have been made. 

(2) [Reserved] 

(b) [Reserved] 

(c) The OL will include appropriate 

provisions with respect to any 

uncompleted items of construction and 

such limitations or conditions as are 

required to assure that operation during 

the period of the completion of such 

items will not endanger public health 

and safety. 

(d) The Commission will issue an OL 

in such form and containing such 

conditions and limitations, including 

technical specifications, as it deems 

necessary and appropriate. 

§ 53.1390 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 § 53.1590. 

§ 53.1396 Duration of operating licenses. 

The Commission will issue an OL 

under this part for the term requested by 

the applicant, not to exceed 40 years 

from the date of issuance, or for the 

estimated useful life of the facility if the 

Commission determines that the 

estimated useful life is less than the 

term requested. 

§ 53.1399 Transfer of an operating license. 

An OL may be transferred under 

§ 53.1570. 

§ 53.1402 Application for renewal. 

The filing of an application for a 

renewed license must be in accordance 

with § 53.1595. 

§ 53.1405 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. 

§ 53.1410 Combined licenses. 

Sections 53.1410 through 53.1461 set 

out the requirements and procedures 

applicable to Commission issuance of 

COLs for commercial nuclear plants 

under this part. 

§ 53.1413 Contents of applications for 

combined licenses; general information. 

An application for a COL must 

include the information required by 

§ 53.1109 and the following 

information: 

(a) Except for an electric utility 

applicant in regard to financial 

assurance required after a Commission 

finding under § 53.1452, the application 

must include information sufficient to 

demonstrate to the Commission the 

financial qualification of the applicant 

to carry out, in accordance with the 

regulations in this chapter, the activities 

for which the permit or license is 

sought. As applicable, the following 

should be provided: 

(1) The applicant must submit 

information that demonstrates that the 

applicant possesses or has reasonable 


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assurance of obtaining the funds 

necessary to cover estimated 

construction costs and related fuel cycle 

costs. The applicant must submit 

estimates of the total construction costs 

of the facility and related fuel cycle 

costs and must indicate the source(s) of 

funds to cover these costs. 

(2) The applicant must submit 

information that demonstrates the 

applicant possesses or has reasonable 

assurance of obtaining the funds 

necessary to cover estimated operation 

costs for the period of the license. The 

applicant must submit estimates for 

total annual operating costs for each of 

the first 5 years of operation of the 

facility. The applicant must also 

indicate the source(s) of funds to cover 

these costs. 

(3) Each application for a COL 

submitted by a newly-formed entity 

organized for the primary purpose of 

constructing and operating a facility 

must also include information 

showing— 

(i) The legal and financial 

relationships the entity has or proposes 

to have with its stockholders or owners; 

and 

(ii) The stockholders’ or owners’ 

financial ability to meet any contractual 

obligation to the entity which they have 

incurred or proposed to incur; and 

(iii) Any other information considered 

necessary by the Commission to enable 

it to determine the applicant’s financial 

qualification. 

(4) The Commission may request an 

established entity or newly formed 

entity to submit additional or more 

detailed information respecting its 

financial arrangements and status of 

funds if the Commission considers this 

information appropriate. This may 

include information regarding a 

licensee’s ability to continue the 

conduct of the activities authorized by 

the license and to decommission the 

facility. 

(b) The application must include 

information in the form of a report, as 

described in subpart G of this part, 

indicating how reasonable assurance 

will be provided that funds will be 

available to decommission the facility. 

§ 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 § 53.1146(a)(1)(iv) through (x). 

(2) Design information. An 

application for a COL for a commercial 

nuclear plant must include the design 

information equivalent to that required 

for a standard design certification as 

defined in § 53.1239(a)(2) through (7), 

(a)(9), and (a)(11) through (27). 

(3) Technical qualifications. A 

description of the technical 

qualifications of the applicant to engage 

in the proposed activities in accordance 

with the regulations in this chapter. 

(4) Integrity assessment program. A 

description of an Integrity Assessment 

Program that addresses the elements 

described in § 53.870. 

(5) Safeguards information. A 

description of the program to protect 

Safeguards Information against 

unauthorized disclosure in accordance 

with the requirements in §§ 73.21 and 

73.22 of this chapter, as applicable. 

(6) Emergency response facility or 

facilities. Description of the locations 

and capabilities to be established for 

command and control, support, and 

coordination of onsite and offsite, as 

applicable, functions during reactor 

accident conditions. 

(7) Role of personnel. (i) A description 

of the completed assessments related to 

the role of personnel in ensuring safe 

operations considering the analyses 

required by § 53.730. These assessments 

must include the following: 

(A) Human factors engineering design 

requirements of § 53.730(a); 

(B) Human system interface design 

requirements of § 53.730(b); 

(C) Concept of operations of 

§ 53.730(c); and 

(D) Functional requirements analysis 

and function allocation of § 53.730(d); 

(ii) A description of the program to be 

used for evaluating and applying 

operating experience as required by 

§ 53.730(e); 

(iii) A staffing plan and supporting 

analyses as required by § 53.730(f). 

(8) Training, examination, and 

proficiency programs. (i) A description 

of the training, examination, and 

proficiency programs required by 

§ 53.730(g); and 

(ii) A description of the training 

programs required by § 53.830. 

(9) Emergency plan. Emergency plans 

complying with the requirements of 

§ 53.855. 

(i) The emergency plan must include, 

as applicable, all emergency plan 

certifications that have been obtained 

from the State, local, and participating 

Tribal governmental agencies with 

emergency planning responsibilities. 

The certifications must state that— 

(A) The proposed emergency plans 

are practicable; 

(B) These agencies are committed to 

participating in any further 

development of the plans, including any 

required field demonstrations; and 

(C) These agencies are committed to 

executing their responsibilities under 

the plans in the event of an emergency. 

(ii) If certifications cannot be obtained 

after sustained, good faith efforts by the 

applicant, then the application must 

contain information, including a utility 

plan, sufficient to show that the 

proposed plans provide reasonable 

assurance that adequate protective 

measures can and will be taken in the 

event of a radiological emergency at the 

site. 

(10) Organization. A description of 

the applicant’s organizational structure, 

allocations of responsibilities and 

authorities, and personnel qualifications 

requirements for operation. 

(11) Maintenance program. A 

description of a maintenance program 

under § 53.715. 

(12) Quality assurance. A description 

of the QAP under § 53.865. 

(13) Radiation protection program. A 

radiation protection program 

description under § 53.850. 

(14) Security program. A physical 

security plan that describes how the 

applicant will comply with § 53.860 

(and 10 CFR part 11, if applicable, 

including the identification and 

description of jobs as required by 

§ 11.11(a) of this chapter, at the 

proposed facility). The plan must list 

tests, inspections, audits, and other 

means to be used to demonstrate 

compliance with the requirements of 10 

CFR parts 11 and 73, if applicable. 

(15) Safeguards contingency plan. A 

safeguards contingency plan in 

accordance with the criteria set forth in 

appendix C to 10 CFR part 73. The 

safeguards contingency plan must 

include plans for dealing with threats, 

thefts, and radiological sabotage, as 

defined in 10 CFR part 73, relating to 

the SNM and nuclear facilities licensed 

under this chapter and in the 

applicant’s possession and control. Each 


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application for this type of license must 

include the information contained in 

the applicant’s safeguards contingency 

plan.1 (Implementing procedures 

required for this plan need not be 

submitted for approval.) 

(16) Security training and 

qualification. A training and 

qualification plan that describes how 

the applicant will demonstrate 

compliance with the criteria set forth in 

§ 73.100 of this chapter or appendix B 

to 10 CFR part 73. 

(17) Cybersecurity plan. A 

cybersecurity plan in accordance with 

the criteria set forth in § 73.54 or 

§ 73.110 of this chapter. 

(18) Security, safeguards and 

cybersecurity plan implementation. A 

description of the implementation of the 

physical security plan, safeguards 

contingency plan, training and 

qualification plan, and cybersecurity 

plan. Each applicant who prepares a 

physical security plan, a safeguards 

contingency plan, a training and 

qualification plan, or a cybersecurity 

plan must protect the plans and other 

related Safeguards Information against 

unauthorized disclosure in accordance 

with the requirements of §§ 73.21 and 

73.22 of this chapter. 

(19) Fire protection program. A 

description of the fire protection 

program under § 53.875. 

(20) Inservice inspection/inservice 

testing program. Descriptions of 

inservice inspection and inservice 

testing programs under § 53.880. 

(21) [Reserved] 

(22) [Reserved] 

(23) General employee training. A 

description of the training program 

required to demonstrate compliance 

with § 53.830 and its implementation. 

(24) Fitness-for-duty program. A 

description of the FFD program under 

part 26 of this chapter and its 

implementation. 

(25) Technical specifications. 

Proposed technical specifications 

prepared in accordance with the 

requirements of § 53.710(a). 

(b) If there are SSCs of the plant for 

which research and development is 

necessary to confirm the adequacy of 

their design, a report which documents 

the resolution of any safety questions 

associated with such SSCs. 

(c) A description of how the 

performance of each safety design 

feature has been demonstrated capable 

of fulfilling functional design criteria 

considering interdependent effects 

through either analysis, appropriate test 

programs, prototype testing, operating 

experience, or a combination thereof, in 

accordance with § 53.440(a). 

(d) If the COL application references 

an early site permit, then the following 

requirements apply: 

(1) The FSAR need not contain 

information or analyses submitted to the 

Commission in connection with the 

early site permit provided that the FSAR 

must either include or incorporate by 

reference the early site permit Site 

Safety Analysis Report and contain, in 

addition to the information and analyses 

otherwise required, information 

sufficient to demonstrate that the design 

of the facility falls within the site 

characteristics and design parameters 

specified in the early site permit. 

(2) If the FSAR does not demonstrate 

that design of the facility falls within 

the site characteristics and design 

parameters, the application must 

include a request for a variance that 

complies with the requirements of 

§§ 53.1188(d) and 53.1437. 

(3) The FSAR must demonstrate that 

all terms and conditions that have been 

included in the early site permit will be 

satisfied by the date of issuance of the 

COL. Any terms or conditions of the 

early site permit that could not be met 

by the time of issuance of the COL must 

be set forth as terms or conditions of the 

COL. 

(4) If the early site permit approves 

complete and integrated emergency 

plans, or major features of emergency 

plans, then the FSAR must include any 

new or additional information that 

updates and corrects the information 

that was provided under § 53.1146(b)(2) 

and discuss whether the new or 

additional information materially 

changes the bases for compliance with 

the applicable requirements. The 

application must identify changes to the 

emergency plans or major features of 

emergency plans that have been 

incorporated into the proposed facility 

emergency plans and that constitute or 

would constitute a change in an 

emergency plan that results in reducing 

the licensee’s capability to perform an 

emergency planning function in the 

event of a radiological emergency. 

(5) If complete and integrated 

emergency plans are approved as part of 

the early site permit, new certifications 

meeting the requirements of paragraph 

(a)(9)(i) of this section are not required. 

(e) If the COL application references 

a standard design approval, then the 

following requirements apply: 

(1) The FSAR need not contain 

information or analyses submitted to the 

Commission in connection with the 

design approval, provided, however, 

that the FSAR must either include or 

incorporate by reference the standard 

design approval FSAR and must 

contain, in addition to the information 

and analyses otherwise required, 

information sufficient to demonstrate 

that the characteristics of the site fall 

within the site parameters specified in 

the design approval. In addition, the 

plant-specific PRA information must 

use the PRA information for the design 

approval and must be updated to 

account for site specific design 

information and any design changes or 

departures. 

(2) The FSAR must demonstrate that 

all terms and conditions that have been 

included in the design approval will be 

satisfied by the date of issuance of the 

COL. 

(f) If the COL application references a 

standard design certification, then the 

following requirements apply: 

(1) The FSAR need not contain 

information or analyses submitted to the 

Commission in connection with the 

standard design certification, provided, 

however, that the FSAR must either 

include or incorporate by reference the 

standard design certification FSAR and 

must contain, in addition to the 

information and analyses otherwise 

required, information sufficient to 

demonstrate that the site characteristics 

fall within the site parameters specified 

in the standard design certification. In 

addition, the plant-specific PRA 

information must use the PRA 

information for the standard design 

certification and must be updated to 

account for site-specific design 

information and any design changes or 

departures. 

(2) The FSAR must demonstrate that 

the interface requirements established 

for the design under § 53.1239(a)(24) 

have been met. 

(3) The FSAR must demonstrate that 

all requirements and restrictions set 

forth in the referenced standard design 

certification rule must be satisfied by 

the date of issuance of the COL. Any 

requirements and restrictions set forth 

in the referenced standard design 

certification rule that could not be 

satisfied by the time of issuance of the 

COL, must be set forth as terms or 

conditions of the COL. 

(g) If the COL application references 

the use of one or more manufactured 

reactors licensed under § 53.1270, then 

the following requirements apply: 

(1) The FSAR need not contain 

information or analyses submitted to the 

Commission in connection with the ML, 

provided, however, that the FSAR must 

either include or incorporate by 

reference the ML FSAR and must 

contain, in addition to the information 

and analyses otherwise required, 

information sufficient to demonstrate 

that the site characteristics fall within 

the site parameters specified in the ML. 


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In addition, the plant-specific PRA 

information must use the PRA 

information for the manufactured 

reactor and must be updated to account 

for site-specific design information and 

any design changes or departures. 

(2) The FSAR must demonstrate that 

the interface requirements established 

for the design have been met. 

(3) The FSAR must demonstrate that 

all terms and conditions that have been 

included in the ML will be satisfied by 

the date of issuance of the COL. Any 

terms or conditions of the ML that could 

not be met by the time of issuance of the 

COL, must be set forth as terms or 

conditions of the COL. 

(h) Each applicant for a COL under 

this part must protect Safeguards 

Information against unauthorized 

disclosure in accordance with the 

requirements in §§ 73.21 and 73.22 of 

this chapter, as applicable. 

1A physical security plan that contains all 

the information required in both § 73.55 or 

§ 73.100 of this chapter and appendix C to 10 

CFR part 73 demonstrates compliance with 

the requirement for a contingency plan. 

§ 53.1419 Contents of applications for 

combined licenses; other application 

content. 

(a) In addition to the FSAR, the 

application must also include the 

following: 

(1) Environmental report. (i) An 

environmental report either in 

accordance with § 51.50(c) of this 

chapter if an LWA under § 53.1130 is 

not requested in conjunction with the 

COL application, or in accordance with 

§§ 51.49 and 51.50(c) of this chapter if 

an LWA is requested in conjunction 

with the COL application; or 

(ii) If the applicant wishes to request 

that an LWA under § 53.1130 be issued 

before issuance of the COL, the 

information otherwise required by 

§ 53.1130, in accordance with either 

§ 2.101(a)(1) through (a)(4), or 

§ 2.101(a)(9) of this chapter; 

(2) Availability controls (if not 

included in the FSAR). A description of 

the controls on plant operations, 

including availability controls, to 

provide reasonable confidence of safe 

operation and that the configurations 

and special treatments for SR SSCs and 

NSRSS SSCs provide the capabilities 

and reliabilities required to satisfy the 

safety criteria of § 53.220, or more 

restrictive alternative criteria adopted 

under § 53.470, if not addressed by 

Technical Specifications under 

§ 53.1416(a)(25); and 

(3) Inspections, tests, analyses, and 

acceptance criteria. The proposed 

inspections, tests, and analyses, 

including those applicable to emergency 

planning, that the licensee must 

perform, and the acceptance criteria that 

are necessary and sufficient to provide 

reasonable assurance that, if the 

inspections, tests, and analyses are 

performed and the acceptance criteria 

met, the facility has been constructed 

and will be operated in conformity with 

the COL, the provisions of the Act, and 

the Commission’s rules and regulations. 

(i) If the application references an 

early site permit with ITAAC, the early 

site permit ITAAC must apply to those 

aspects of the COL which are approved 

in the early site permit. 

(ii) If the application references a 

standard design certification, the ITAAC 

contained in the certified design must 

apply to those portions of the facility 

design which are approved in the 

standard design certification. 

(iii) If the application references an 

ML, the ITAAC contained in the ML 

must apply to those portions of the 

facility design which are approved in 

the ML. 

(iv) If the application references an 

early site permit with ITAAC, a 

standard design certification, an ML, or 

combination thereof, the application 

may include a notification that a 

required inspection, test, or analysis in 

the ITAAC has been successfully 

completed and that the corresponding 

acceptance criterion has been met. The 

Federal Register notification required 

by § 53.1422 of this chapter must 

indicate that the application includes 

this notification. 

(b) [Reserved] 

§ 53.1422 Review of applications. 

(a) Standards for review of 

applications. Applications filed under 

this part will be reviewed according to 

the standards set out in this part and 10 

CFR parts 20, 51, 73, and 140. 

(b) Administrative review of 

applications; hearings. A proceeding on 

a COL is subject to all applicable 

procedural requirements contained in 

10 CFR part 2, including the 

requirements for docketing (§ 2.101 of 

this chapter) and issuance of a notice of 

hearing (§ 2.104 of this chapter). If an 

applicant requests a Commission 

finding on certain ITAAC with the 

issuance of the COL, then those ITAAC 

will be identified in the notice of 

hearing. All hearings on COLs are 

governed by the procedures contained 

in 10 CFR part 2. 

§ 53.1425 Finality of referenced NRC 

approvals. 

If the application for a COL under this 

part references an early site permit, 

standard design certification rule, 

standard design approval, or ML, issued 

under this part, the scope and nature of 

matters resolved for the application and 

any COL issued are governed by the 

relevant provisions addressing finality, 

including §§ 53.1188, 53.1221, 53.1263, 

and 53.1288. 

§ 53.1431 Referral to the Advisory 

Committee on Reactor Safeguards. 

The Commission must refer a copy of 

the application to the ACRS. The ACRS 

must report on those portions of the 

application that concern safety and 

must apply the standards referenced in 

§ 53.1422, in accordance with the 

finality provisions in § 53.1425. 

§ 53.1434 Authorization to conduct limited 

work authorization activities. 

(a) If the application for a COL under 

this part does not reference an early site 

permit which authorizes the holder to 

perform the activities under 

§ 53.1130(b), the applicant may not 

perform those activities without 

obtaining the separate authorization 

required by § 53.1130(a). Authorization 

may be granted only after the presiding 

officer in the proceeding on the 

application has made the findings and 

determination required by 

§ 53.1130(b)(1)(ii) and (b)(1)(iv), and the 

Director, Office of Nuclear Reactor 

Regulation makes the determination 

required by § 53.1130(b)(1)(iii). 

(b) If, after an applicant has performed 

the activities permitted by a LWA 

issued under § 53.1130, the application 

for the COL is withdrawn or denied, 

then the applicant must implement the 

approved site redress plan. 

§ 53.1437 Exemptions, departures, and 

variances. 

(a) An applicant for a COL, or any 

amendment to a COL, may include in 

the application a request for an 

exemption from one or more of the 

Commission’s regulations. 

(1) If the request is for an exemption 

from any part of a referenced standard 

design certification rule, the 

Commission may grant the request if it 

determines that the exemption complies 

with any exemption provisions of the 

referenced standard design certification 

rule, or with § 53.1263 if there are no 

applicable exemption provisions in the 

referenced standard design certification 

rule. 

(2) For all other requests for 

exemptions, the Commission may grant 

a request if it determines that the 

exemption complies with § 53.080. 

(b) An applicant for a COL who has 

filed an application referencing an early 

site permit issued under § 53.1158 may 

include in the application a request for 

a variance from one or more site 

characteristics, design parameters, or 


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terms and conditions of the permit, or 

from the Site Safety Analysis Report. In 

determining whether to grant the 

variance, the Commission must apply 

the same technically relevant criteria as 

were applicable to the application for 

the original or renewed site permit. 

Once a COL referencing an early site 

permit is issued, variances from the 

early site permit will not be granted for 

that CP or COL. 

(c) An applicant for a COL who has 

filed an application referencing use of a 

manufactured reactor may include in 

the application a request for a departure 

from one or more design characteristics, 

site parameters, terms and conditions, 

or approved design of the manufactured 

reactor under the ML issued under 

§ 53.1287. The Commission may grant 

such a request only if it determines that 

the departure will comply with the 

requirements of § 53.080, and that the 

special circumstances outweigh any 

decrease in safety that may result from 

the reduction in standardization caused 

by the departure. 

(d) Issuance of a variance under 

paragraph (b) of this section or a 

departure under paragraph (c) of this 

section is subject to litigation during the 

COL proceeding in the same manner as 

other issues material to that proceeding. 

§ 53.1440 Issuance of combined licenses. 

(a)(1) After conducting a hearing 

under § 53.1422(b) and receiving the 

report submitted by the ACRS, the 

Commission may issue a COL if the 

Commission finds that— 

(i) The applicable standards and 

requirements of the Act and the 

Commission’s regulations have been 

met; 

(ii) Any required notifications to other 

agencies or bodies have been duly 

made; 

(iii) There is reasonable assurance that 

the facility will be constructed and will 

operate in conformity with the license, 

the provisions of the Act, and the 

Commission’s regulations; 

(iv) The applicant is technically and 

financially qualified to engage in the 

activities authorized; however, no 

finding of financial qualification is 

necessary for an electric utility 

applicant for a COL; 

(v) Issuance of the license will not be 

inimical to the common defense and 

security or to the health and safety of 

the public; and 

(vi) The findings required by subpart 

A of 10 CFR part 51 have been made. 

(2) The Commission may also find, at 

the time it issues the COL, that certain 

acceptance criteria in one or more of the 

ITAAC in a referenced early site permit, 

standard design certification, or ML 

have been met. This finding will finally 

resolve that those acceptance criteria 

have been met, those acceptance criteria 

will be deemed to be excluded from the 

COL, and findings under § 53.1452(g) 

with respect to those acceptance criteria 

are unnecessary. 

(b) The Commission must identify 

within the COL the inspections, tests, 

and analyses, including those applicable 

to emergency planning, that the licensee 

must perform, and the acceptance 

criteria that, if met, are necessary and 

sufficient to provide reasonable 

assurance that the facility has been 

constructed and will be operated in 

conformity with the license, the 

provisions of the Act, and the 

Commission’s rules and regulations. 

(c) A COL must contain the terms and 

conditions, including technical 

specifications, as the Commission 

deems necessary and appropriate. 

§ 53.1443 Finality of combined licenses. 

(a) After issuance of a COL, the 

Commission may not modify, add, or 

delete any term or condition of the COL, 

the design of the facility, the ITAAC 

contained in the license that are not 

derived from a referenced standard 

design certification or ML, except under 

the provisions of § 53.1452 or § 53.1590. 

(b) If the COL does not reference a 

standard design certification or use of a 

manufactured reactor under an ML 

issued under § 53.1287, then a licensee 

may make changes in the facility as 

described in the FSAR (as updated) and 

make changes in the procedures as 

described in the FSAR (as updated) 

under the applicable change processes 

in § 53.1550. 

(c) If the COL references a certified 

design, then— 

(1) Changes to or departures from 

information within the scope of the 

referenced standard design certification 

rule are subject to the applicable change 

processes in that rule; and 

(2) Changes that are not within the 

scope of the referenced standard design 

certification rule are subject to the 

applicable change processes in subpart 

I of this part, unless they also involve 

changes to or noncompliance with 

information within the scope of the 

referenced standard design certification 

rule. In these cases, the applicable 

provisions of this section and the 

standard design certification rule apply. 

(d) If the COL references use of a 

manufactured reactor under an ML 

issued under this part, then— 

(1) Changes to or departures from 

information within the scope of the 

manufactured reactor’s design are 

subject to the change processes in 

§ 53.1288; and 

(2) Changes that are not within the 

scope of the manufactured reactor’s 

design are subject to the applicable 

change processes in subpart I. 

(e) The Commission may issue and 

make immediately effective any 

amendment to a COL upon a 

determination by the Commission that 

the amendment involves no significant 

hazards consideration, notwithstanding 

the pendency before the Commission of 

a request for a hearing from any person. 

The amendment may be issued and 

made immediately effective in advance 

of the holding and completion of any 

required hearing. The amendment will 

be processed under the procedures 

specified in § 53.1515. 

(f) Any modification to, addition to, or 

deletion from the terms and conditions 

of a COL, including any modification to, 

addition to, or deletion from the 

inspections, tests, and analyses, or 

related acceptance criteria contained in 

the license is a proposed amendment to 

the license. There must be an 

opportunity for a hearing on the 

amendment. 

§ 53.1449 Inspection during construction. 

(a) Licensee schedule for inspections, 

tests, or analyses. The licensee must 

submit to the NRC, no later than 1 year 

after issuance of the COL or at the start 

of construction as defined at § 53.020, 

whichever is later, its schedule for 

completing the inspections, tests, or 

analyses in the ITAAC. The licensee 

must submit updates to the ITAAC 

schedules every 6 months thereafter 

and, within 1 year of its scheduled date 

for initial loading of fuel (or, for a fueled 

manufactured reactor, within 1 year of 

its scheduled date for initiating the 

physical removal of any one of the 

independent physical mechanisms to 

prevent criticality required under 

§ 53.620(d)(1)), the licensee must submit 

updates to the ITAAC schedule every 30 

days until the final notification is 

provided to the NRC under paragraph 

(c)(1) of this section. 

(b) Licensee and applicant conduct of 

activities subject to ITAAC. With respect 

to activities subject to an ITAAC, an 

applicant for a COL may proceed at its 

own risk with design and procurement 

activities, and a licensee may proceed at 

its own risk with design, procurement, 

construction, and preoperational 

activities, even though the NRC may not 

have found that any one of the 

prescribed acceptance criteria are met. 

(c) Licensee notifications. (1) ITAAC 

closure notification. The licensee must 

notify the NRC that prescribed 

inspections, tests, and analyses have 

been performed and that the prescribed 

acceptance criteria are met. The 


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notification must contain sufficient 

information to demonstrate that the 

prescribed inspections, test, and 

analyses have been performed and that 

the prescribed acceptance criteria are 

met. 

(2) ITAAC post-closure notifications. 

Following the licensee’s ITAAC closure 

notifications under paragraph (c)(1) of 

this section until the Commission makes 

the finding under § 53.1452(g), the 

licensee must notify the NRC, in a 

timely manner, of new information that 

materially alters the basis for 

determining that either inspections, 

tests, or analyses were performed as 

required, or that acceptance criteria are 

met. The notification must contain 

sufficient information to demonstrate 

that, notwithstanding the new 

information, the prescribed inspections, 

tests, and analyses have been performed 

as required, and the prescribed 

acceptance criteria are met. 

(3) Uncompleted ITAAC notification. 

If the licensee has not provided, by the 

date 225 days before the scheduled date 

for initial loading of fuel (or, for a fueled 

manufactured reactor, by the date 225 

days before the scheduled date for 

initiating the physical removal of any 

one of the independent physical 

mechanisms to prevent criticality 

required under § 53.620(d)(1)), the 

notification required by paragraph (c)(1) 

of this section for all ITAAC, then the 

licensee must notify the NRC that the 

prescribed inspections, tests, or analyses 

for all uncompleted ITAAC will be 

performed and that the prescribed 

acceptance criteria will be met prior to 

operation. The notification must be 

provided no later than the date 225 days 

before the scheduled date for initial 

loading of fuel (or, for a fueled 

manufactured reactor, no later than the 

date 225 days before the scheduled date 

for initiating the physical removal of 

any one of the independent physical 

mechanisms to prevent criticality 

required under § 53.620(d)(1)), and must 

provide sufficient information to 

demonstrate that the prescribed 

inspections, tests, or analyses will be 

performed and the prescribed 

acceptance criteria for the uncompleted 

ITAAC will be met, including, but not 

limited to, a description of the specific 

procedures and analytical methods to be 

used for performing the prescribed 

inspections, tests, and analyses and 

determining that the prescribed 

acceptance criteria are met. 

(4) All ITAAC complete notification. 

The licensee must notify the NRC that 

all ITAAC are complete. 

(d) Licensee determination of 

noncompliance with ITAAC. (1) In the 

event that an activity is subject to an 

ITAAC derived from a referenced 

standard design certification and the 

licensee has not demonstrated that the 

prescribed acceptance criteria are met, 

the licensee may take corrective actions 

to successfully complete that ITAAC or 

request an exemption from the standard 

design certification ITAAC, as 

applicable. A request for an exemption 

must also be accompanied by a request 

for a license amendment under subpart 

I. 

(2) In the event that an activity is 

subject to an ITAAC not derived from a 

referenced standard design certification 

and the licensee has not demonstrated 

that the prescribed acceptance criteria 

are met, the licensee may take corrective 

actions to successfully complete that 

ITAAC or request a license amendment 

under subpart I. 

(e) NRC inspection, publication of 

notices, and availability of licensee 

notifications. The NRC must ensure that 

the prescribed inspections, tests, and 

analyses in the ITAAC are performed. 

(1) At appropriate intervals until the 

last date for submission of requests for 

hearing under § 53.1452, the NRC must 

publish notices in the Federal Register 

of the NRC staff’s determination of the 

successful completion of inspections, 

tests, and analyses. 

(2) The NRC must make publicly 

available the licensee notifications 

under paragraph (c) of this section. The 

NRC must, no later than the date of 

publication of the notice of intended 

operation required by § 53.1452(a), 

make publicly available those licensee 

notifications under paragraph (c) of this 

section that have been submitted to the 

NRC at least 7 days before that notice. 

§ 53.1452 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 

§ 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 § 2.309 of this chapter. 

If the Commission grants the request, 

the Commission, acting as the presiding 

officer, must determine whether during 

a period of interim operation there will 

be reasonable assurance of adequate 

protection to the public health and 

safety. The Commission’s determination 

must consider the petitioner’s prima 

facie showing and any answers thereto. 

If the Commission determines there is 

such reasonable assurance, it must 

allow operation during an interim 

period under the COL. 

(d) The Commission, in its discretion, 

must determine appropriate hearing 

procedures, whether informal or formal 

adjudicatory, for any hearing under 

paragraph (a) of this section, and must 

state its reasons therefore. 

(e) The Commission must, to the 

maximum possible extent, render a 

decision on issues raised by the hearing 

request within 180 days of the 

publication of the notice provided by 

paragraph (a) of this section or by the 

anticipated date for initial loading of 

fuel into the reactor, whichever is later. 

(f) A petition to modify the terms and 

conditions of the COL will be processed 

as a request for action under § 2.206 of 

this chapter. The petitioner must file the 

petition with the Secretary of the 

Commission. Before the licensed 

activity allegedly affected by the 

petition (fuel loading, low power 

testing, etc.) commences, the 

Commission must determine whether 

any immediate action is required. If the 

petition is granted, then an appropriate 

order will be issued. Fuel loading and 

operation under the COL will not be 

affected by the granting of the petition 

unless the order is made immediately 

effective. 

(g) The licensee must not operate the 

facility until the Commission makes a 

finding that the acceptance criteria in 

the COL are met, except for those 

acceptance criteria that the Commission 

found were met under § 53.1440(a)(2). If 

the COL is for a modular design, each 


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reactor unit may require a separate 

finding as construction proceeds. 

(h) After the Commission has made 

the finding in paragraph (g) of this 

section, the ITAAC do not, by virtue of 

their inclusion in the COL, constitute 

regulatory requirements either for 

licensees or for renewal of the license; 

except for the specific ITAAC for which 

the Commission has granted a hearing 

under paragraph (a) of this section, all 

ITAAC expire upon final Commission 

action in the proceeding. However, 

subsequent changes to the facility or 

procedures described in the FSAR (as 

updated) must comply with the 

requirements in § 53.1443(e) or (f), as 

applicable. 

1For licensees installing fueled 

manufactured reactors under a COL, the COL 

holder must instead notify the NRC of its 

scheduled date for initiating the physical 

removal of any one of the independent 

physical mechanisms to prevent criticality 

required under § 53.620(d)(1) no later than 

270 days before the scheduled date and must 

notify the NRC of updates to its schedule 

every 30 days thereafter. 2For licensees installing fueled 

manufactured reactors under a COL, the 

Commission must instead publish notice of 

intended operation in the Federal Register 

not less than 180 days before the date 

scheduled for initiating the physical removal 

of any one of the independent physical 

mechanisms to prevent criticality required 

under § 53.620(d)(1). 

§ 53.1455 Duration of combined license. 

A COL is issued for a specified period 

not to exceed 40 years from the date on 

which the Commission makes a finding 

that acceptance criteria are met under 

§ 53.1452(g) or allowing operation 

during an interim period under the COL 

under § 53.1452(c). 

§ 53.1456 Transfer of a combined license. 

A COL may be transferred under 

§ 53.1570. 

§ 53.1458 Application for renewal. 

The filing of an application for a 

renewed license must be in accordance 

with § 53.1595. 

§ 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. 

§ 53.1470 Standardization of commercial 

nuclear plant designs: licenses to construct 

and operate nuclear power reactors of 

identical design at multiple sites. 

(a) Except as otherwise specified in 

this section, the provisions of this 

section apply to CP, OL, and COL 

applications for commercial nuclear 

plants of identical design (the ‘‘common 

design’’) under this part. 

(b) Each application for a CP, OL, or 

COL submitted pursuant to this section 

must be submitted as specified in 

§§ 53.1300, 53.1360, or 53.1410, 

respectively, and § 2.101 of this chapter. 

Each application must state that the 

applicant wishes to construct a facility 

identical to a facility proposed for one 

or more sites other than the applicant’s 

(the ‘‘common design’’), and the 

applicant wishes to have the application 

considered under this section. Each 

application must list each of the other 

applications to be treated together under 

this section. 

(c) Each application must include the 

information required by the applicable 

sections of this subpart, provided 

however, that the application must 

identify the common design, and, if 

applicable, reference a standard design 

certification or standard design approval 

under this part, or the use of a reactor 

manufactured under this part. The 

FSAR for each application must either 

incorporate by reference or include the 

final safety analysis of the common 

design, including, if applicable, the 

FSAR for the referenced standard design 

certification, standard design approval, 

or the manufactured reactor. 

(d) Each application submitted 

pursuant to this section must contain an 

environmental report under 

§§ 53.1312(a)(1), 53.1372(a), or 

53.1419(a)(1), as applicable, that 

complies with the applicable provisions 

of 10 CFR part 51, provided, however, 

that the application may incorporate by 

reference a single environmental report 

on the environmental impacts of the 

common design that are applicable to 

each site. 

(e) Upon a determination that each 

application is acceptable for docketing 

under § 2.101 of this chapter, each 

application will be docketed and a 

notice of docketing for each application 

will be published in the Federal 

Register, under § 2.104 of this chapter, 

provided, however, that the notice must 

state that the application will be 

processed under the provisions of this 

section and subpart D of 10 CFR part 2. 

At the discretion of the Commission, a 

single notice of docketing for multiple 

applications may be published in the 

Federal Register. 

(f) The NRC must prepare an 

environmental assessment or draft and 

final environmental impact statements 

for each of the applications under 10 

CFR part 51. Scoping under §§ 51.28 

and 51.29 of this chapter for each of the 

license applications may be conducted 

simultaneously and joint scoping may 

be conducted with respect to the 

environmental issues relevant to the 

common design. If the applications 

reference a standard design certification, 

then the environmental assessment or 

environmental impact statement for 

each of the applications must 

incorporate by reference the standard 

design certification environmental 

assessment. If the applications do not 

reference a standard design certification, 

then the NRC must prepare 

environmental assessments or draft and 

final supplemental environmental 

impact statements which address severe 

accident mitigation design alternatives 

for the common design, which must be 

incorporated by reference into the 

environmental assessment or 

environmental impact statement 

prepared for each application. Scoping 

under §§ 51.28 and 51.29 of this chapter 

for the supplemental environmental 

impact statement may be conducted 

simultaneously and may be part of the 

scoping for each of the applications. 

(g) The ACRS must report on each of 

the applications as required by the 

applicable sections of this subpart. Each 

report must be limited to those safety 

matters for each application that are not 

relevant to the common design. In 

addition, the ACRS must separately 

report on the safety of the common 

design, provided, however, that the 

report need not address the safety of a 

referenced standard design certification 

or reactor manufactured under this part. 

(h) The Commission must designate a 

presiding officer to conduct the 

proceeding with respect to the health 

and safety, common defense and 

security, and environmental matters 

relating to the common design and 

affecting at least two applications. The 

hearing will be governed by the 

applicable provisions of subparts A, C, 

G, L, N, and O of 10 CFR part 2 relating 

to applications for CPs, OLs, and COLs. 

The presiding officer must issue a 

partial initial decision on the common 

design. 

(i) If the design for the power 

reactor(s) proposed in a particular 


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application is not identical to the others, 

that application may not be processed 

under this section and subpart D of 10 

CFR part 2. 

(j) As used in this section, the design 

of a nuclear power reactor included in 

a single referenced Safety Analysis 

Report means the design of those SSCs 

important to radiological health and 

safety and the common defense and 

security. 

Subpart I—Maintaining and Revising 

Licensing-Basis Information 

§ 53.1500 Licensing-basis information. 

This subpart provides the 

requirements for each holder of a 

license for a commercial nuclear plant 

licensed under this part to maintain 

licensing-basis information as defined 

in § 53.020; evaluate changes to site 

characteristics, plant design features, 

and programmatic controls to determine 

needed approvals and revisions; and 

submit appropriate updates to the U.S. 

Nuclear Regulatory Commission (NRC). 

§ 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 § 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 

§ 50.160(b)(1)(iv)(B) apply, then the 

NRC will base its finding on a review of 

the Federal Emergency Management 

Agency findings and determinations as 

to whether State, participating Tribal 

and local emergency plans are adequate 

and capable of being implemented, and 

on the NRC assessment as to whether 

the licensee’s emergency plans are 

adequate and capable of being 

implemented. Nothing in this paragraph 

must be construed as limiting the 

authority of the Commission to take 

action under any other regulation or 

authority of the Commission or at any 

time other than that specified in this 

paragraph. 

§ 53.1505 Changes to licensing-basis 

information requiring prior NRC approval. 

(a) Sections 53.1510 through 53.1520 

provide the process for a licensee to 

request and the NRC to issue 

amendments to licenses, including any 

conditions contained therein, technical 

specifications or other attachments to a 

license, and any orders issued by the 

NRC modifying a license. Sections 

53.1525 and 53.1530 govern proposed 

changes to a commercial nuclear plant 

referencing a certified design or 

manufacturing license (ML). 

(b) A licensee may propose changing 

licensing-basis information established 

by NRC regulations by requesting an 

exemption in accordance with § 53.080. 

§ 53.1510 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 § 53.040, that fully 

describes the changes desired and, 

following as far as applicable, the form 

prescribed for original applications. 

Applications for amendments involving 

changes to plant structures, systems, 

and components (SSCs), programmatic 

controls, or the role of plant personnel 

must include an assessment of the 

changes in relation to the safety 

requirements in subpart B of this part 

and the analyses requirements of 

§ 53.450 as applicable, an analysis of 

whether the amendment involves no 

significant hazards consideration using 

the standards in § 53.1520, and a 

consideration of environmental factors. 

§ 53.1515 Public notices; State 

consultation. 

The Commission will use the 

following procedures for an application 

requesting an amendment to an OL or 

COL issued under this part. 

(a) Public notices. (1)(i) The 

Commission may publish in the Federal 

Register under § 2.105 of this chapter an 

individual notice of proposed action for 

an amendment for which it makes a 

proposed determination that no 

significant hazards consideration is 

involved, or, at least once every 30 days, 

publish a periodic Federal Register 

notice of proposed actions, which 

identifies each amendment issued and 

each amendment proposed to be issued 

since the last such periodic notice, or it 

may publish both such notices. 

(ii) For each amendment proposed to 

be issued, the notice will 

(A) Contain the staff’s proposed 

determination under the standards in 

§ 53.1520; 

(B) Provide a brief description of the 

amendment and of the facility involved; 

(C) Solicit public comments on the 

proposed determination; and 

(D) Provide for a 30-day comment 

period. 

(iii) The comment period will begin 

on the day after the date of the 

publication of the first notice, and, 

normally, the amendment will not be 

granted until after this comment period 

expires. 


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(2) The Commission may inform the 

public about the final disposition of an 

amendment request for which it has 

made a proposed determination of no 

significant hazards consideration either 

by issuing an individual notice of 

issuance under § 2.106 of this chapter or 

by publishing such a notice in its 

periodic system of Federal Register 

notices. In either event, it will not make 

and will not publish a final 

determination of no significant hazards 

consideration unless it receives a 

request for a hearing on that amendment 

request. 

(3) Where the Commission makes a 

final determination that no significant 

hazards consideration is involved and 

that the amendment should be issued, 

the amendment will be effective on 

issuance, even if adverse public 

comments have been received and even 

if an interested person meeting the 

provisions for intervention called for in 

§ 2.309 of this chapter has filed a 

request for a hearing. The Commission 

need hold any required hearing only 

after it issues an amendment, unless it 

determines that a significant hazards 

consideration is involved, in which case 

the Commission will provide an 

opportunity for a prior hearing. 

(4) Where the Commission finds that 

an emergency situation exists, in that 

failure to act in a timely way would 

result in derating or shutdown of a 

commercial nuclear reactor, or in 

prevention of either resumption of 

operation or of increase in power output 

up to the plant’s licensed power level, 

it may issue a license amendment 

involving no significant hazards 

consideration without prior notice and 

opportunity for a hearing or for public 

comment. In such a situation, the 

Commission will not publish a notice of 

proposed determination on no 

significant hazards consideration but 

will publish a notice of issuance under 

§ 2.106 of this chapter providing for 

opportunity for a hearing and for public 

comment after issuance. The 

Commission expects its licensees to 

apply for license amendments in a 

timely fashion. It will decline to 

dispense with notice and comment on 

the determination of no significant 

hazards consideration if it determines 

that the licensee has abused the 

emergency provision by failing to make 

timely application for the amendment 

and thus itself creating the emergency. 

Whenever an emergency situation 

exists, a licensee requesting an 

amendment must explain why this 

emergency situation occurred and why 

it could not avoid this situation, and the 

Commission will assess the licensee’s 

reasons for failing to file an application 

sufficiently in advance of that event. 

(5) Where the Commission finds that 

exigent circumstances exist, in that a 

licensee and the Commission must act 

quickly and that time does not permit 

the Commission to publish a Federal 

Register notice allowing 30 days for 

prior public comment, and it also 

determines that the amendment 

involves no significant hazards 

considerations, it— 

(i)(A) Will either issue a Federal 

Register notice providing notice of an 

opportunity for hearing and allowing at 

least 2 weeks from the date of the notice 

for prior public comment; or 

(B) Will use local media to provide 

reasonable notice to the public in the 

area surrounding a licensee’s facility of 

the licensee’s amendment and of its 

proposed determination as described in 

paragraph (a)(1) of this section, 

consulting with the licensee on the 

proposed media release and on the 

geographical area of its coverage; 

(ii) Will provide for a reasonable 

opportunity for the public to comment, 

using its best efforts to make available 

to the public whatever means of 

communication it can for the public to 

respond quickly, and, in the case of 

telephone comments, have these 

comments recorded or transcribed, as 

necessary and appropriate; 

(iii) When it has issued a local media 

release, may inform the licensee of the 

public’s comments, as necessary and 

appropriate; 

(iv) Will publish a notice of issuance 

under § 2.106 of this chapter; 

(v) Will provide a hearing after 

issuance, if one has been requested by 

a person who satisfies the provisions for 

intervention specified in § 2.309 of this 

chapter; and 

(vi) Will require the licensee to 

explain the exigency and why the 

licensee cannot avoid it and use its 

normal public notice and comment 

procedures in paragraph (a)(1) of this 

section if it determines that the licensee 

has failed to use its best efforts to make 

a timely application for the amendment 

in order to create the exigency and to 

take advantage of this procedure. 

(6) Where the Commission finds that 

significant hazards considerations are 

involved, it will issue a Federal Register 

notice providing an opportunity for a 

prior hearing even in an emergency 

situation, unless it finds an imminent 

danger to the health or safety of the 

public, in which case it will issue an 

appropriate order or rule under 10 CFR 

part 2. 

(b) State consultation. (1) At the time 

a licensee requests an amendment, it 

must notify the State in which its 

facility is located of its request by 

providing that State with a copy of its 

application and its reasoned analysis 

about no significant hazards 

considerations and indicate on the 

application that it has done so. 

(2) The Commission will advise the 

State of its proposed determination 

about no significant hazards 

consideration normally by sending it a 

copy of the Federal Register notice. 

(3) The Commission will make the 

names of the Project Manager or other 

NRC personnel it designated to consult 

with the State available to the State 

official designated to consult about its 

proposed determination. The 

Commission will consider any 

comments of that State official. If it does 

not hear from the State in a timely 

manner, it will consider that the State 

has no interest in its determination; 

nonetheless, to ensure that the State is 

aware of the application, before it issues 

the amendment, it will make a good 

faith effort to communicate directly 

with that official. (Inability to consult 

with a responsible State official 

following good faith attempts will not 

prevent the Commission from making 

effective a license amendment involving 

no significant hazards consideration.) 

(4) The Commission will make a good 

faith attempt to consult with the State 

before it issues a license amendment 

involving no significant hazards 

consideration. If, however, it does not 

have time to use its normal consultation 

procedures because of an emergency 

situation, it will attempt to 

communicate directly with the 

appropriate State official. (Inability to 

consult with a responsible State official 

following good faith attempts will not 

prevent the Commission from making 

effective a license amendment involving 

no significant hazards consideration, if 

the Commission deems it necessary in 

an emergency situation.) 

(5) After the Commission issues the 

requested amendment, it will send a 

copy of its determination to the State. 

(c) Caveats about State consultation. 

(1) The State consultation procedures in 

paragraph (b) of this section do not give 

the State a right— 

(i) To veto the Commission’s 

proposed or final determination; 

(ii) To a hearing on the determination 

before the amendment becomes 

effective; or 

(iii) To insist upon a postponement of 

the determination or upon issuance of 

the amendment. 

(2) These procedures do not alter 

present provisions of law that reserve to 

the Commission exclusive responsibility 

for setting and enforcing radiological 


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health and safety requirements for 

commercial nuclear plants. 

§ 53.1520 Issuance of amendment. 

(a) In determining whether an 

amendment to a license will be issued 

to the applicant, the Commission will be 

guided by the considerations which 

govern the issuance of initial licenses to 

the extent applicable and appropriate. If 

the application is for amendment of an 

OL or COL and involves the material 

alteration of a commercial nuclear plant, 

a construction permit (CP) will be 

issued before the issuance of the 

amendment to the license, provided 

however, that if the application involves 

a material alteration to a manufactured 

reactor under this part before its 

installation at a site, or a COL before the 

date that the Commission makes the 

finding under § 53.1452(g), no 

application for or issuance of a CP is 

required. If the amendment involves a 

significant hazards consideration, the 

Commission will give notice of its 

proposed action— 

(1) Under § 2.105 of this chapter 

before acting thereon; and 

(2) As soon as practicable after the 

application has been docketed. 

(b) The Commission will be 

particularly sensitive to a license 

amendment request that involves 

irreversible consequences (such as one 

that permits a significant increase in the 

amount of effluents or radiation emitted 

by a commercial nuclear plant). 

(c) The Commission may make a final 

determination, under the procedures in 

§ 53.1515, that a proposed amendment 

to an OL or a COL for a commercial 

nuclear plant under this part involves 

no significant hazards consideration, if 

operation of the plant in accordance 

with the proposed amendment would 

not— 

(1) Involve a significant increase in 

the probability or consequences of an 

accident previously evaluated; or 

(2) Create the possibility of a new or 

different kind of an accident from any 

accident previously evaluated; or 

(3) Involve a significant reduction in 

a margin of safety. 

§ 53.1525 Revising certification 

information within a design certification 

rule. 

(a) A holder of an OL or COL who 

references a design certification rule 

issued under this part must request an 

exemption if proposing to change one or 

more elements of the certification 

information. The Commission may grant 

such a request only if it determines that 

the exemption will comply with the 

requirements of § 53.080 and that the 

special circumstances outweigh any 

decrease in safety that may result from 

the reduction in standardization caused 

by the departure. 

(b) The request for an exemption must 

be included with any associated license 

amendment request, which must be 

requested and processed in accordance 

with §§ 53.1510, 53.1515, and 53.1520. 

(c) Licensees must evaluate changes to 

the design as described in the Final 

Safety Analysis Report (FSAR) not 

involving changes to the certification 

information using the criteria in 

§ 53.1550. 

§ 53.1530 Revising design information 

within a manufacturing license. 

(a) The holder of an ML may not make 

changes to the design of the 

manufactured reactor authorized to be 

manufactured without obtaining an 

amendment pursuant to § 53.1510 and, 

as applicable, § 53.1520. 

(b) The holder of a COL under this 

part who references or uses a 

manufactured reactor under this part 

must request approval for any proposed 

departure from the design 

characteristics, site parameters, terms 

and conditions, or approved design of 

the manufactured reactor. The 

application for such departures must be 

submitted and processed in accordance 

with §§ 53.1510, 53.1515, and 53.1520. 

In those cases where an ML references 

a design certification rule, the 

amendment application from the holder 

of the COL must also request an 

exemption from the design certification 

rule under § 53.1525 if one or more 

elements of the certification information 

are adversely affected by the proposed 

change. The holder of the COL must 

evaluate changes to the commercial 

nuclear plant as described in the FSAR 

but outside of the scope of the 

referenced ML using the criteria in 

§ 53.1550. 

§ 53.1535 Amendments during 

construction. 

(a) The holder of a CP or limited work 

authorization (LWA) under this part 

may request an amendment to the CP or 

LWA in order to gain Commission 

approval of the safety of selected design 

features or specifications, including 

proposed departures from a design 

certification rule or ML. Amendments to 

CPs or LWAs under this part must be 

requested and processed under 

§§ 53.1510 and 53.1520. 

(b) The holder of a COL under this 

part for which the NRC has not yet 

made a finding in accordance with 

§ 53.1452(g) must request amendments 

required by § 53.1525 or § 53.1550 no 

later than 45 days from the date the 

licensee begins the construction of the 

SSCs to implement the change or 

departure requiring NRC approval. The 

licensee proceeds with such changes at 

its own risk recognizing that there is a 

possibility that the amendment will not 

be granted. 

§ 53.1540 Updating licensing-basis 

information and determining the need for 

NRC approval. 

(a) Sections 53.1545 through 53.1565 

provide the process for a holder of an 

OL or COL to modify licensing-basis 

information and to evaluate potential 

changes to its facilities, procedures, 

programs, and organizations to 

determine if NRC approval is required. 

(b) Definitions for the purposes of 

§§ 53.1545 through 53.1565— 

Change means a modification or 

addition to, or removal from, the 

commercial nuclear plant or procedures 

that affects a design feature or related 

functional design criteria, method of 

performing or controlling the functions 

of design features, or an evaluation that 

demonstrates that intended functions 

will be accomplished. 

Departure from a method of 

evaluation described in the Final Safety 

Analysis Report (FSAR) (as updated) 

used in establishing the functional 

design criteria for safety-related 

structures, systems, or components or in 

the safety analyses means— 

(1) Changing any of the elements of 

the method described in the FSAR (as 

updated) unless the results of the 

analysis are conservative or essentially 

the same; or 

(2) Changing from a method described 

in the FSAR to another method unless 

that method has been approved by NRC 

for the intended application. 

Facility as described in the FSAR (as 

updated) means— 

(1) The SSCs that are described in the 

FSAR (as updated), 

(2) The design and performance 

requirements for such SSCs described in 

the FSAR (as updated), and 

(3) The evaluations or methods of 

evaluation included in the FSAR (as 

updated) for such SSCs which 

demonstrate that their intended 

function(s) will be accomplished. 

Final Safety Analysis Report (as 

updated) means the FSAR submitted 

under § 53.1369 or § 53.1416, as 

amended and supplemented, and as 

updated under § 53.1545, as applicable. 

Procedures as described in the Final 

Safety Analysis Report (as updated) 

means those procedures that contain 

information described in the FSAR (as 

updated) such as how SSCs are operated 

and controlled (including assumed 

operator actions and response times). 


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§ 53.1545 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 

§ 53.1452(g) must update the FSAR 

originally submitted as part of the 

application for the license every 24 

months or more frequently to assure that 

the information included in the report 

contains the latest information 

developed. The submittal must include 

the effects on the content of the FSAR 

of— 

(1) Changes made to the facility or 

procedures as described in the FSAR; 

(2) Safety analyses and evaluations 

performed by the licensee either in 

support of approved license 

amendments or in support of 

conclusions that changes did not require 

a license amendment under § 53.1550; 

(3) Updates to the probabilistic risk 

assessments required under § 53.450; 

(4) The cumulative effects of the 

changes to the facility or procedures on 

the margins to the safety criteria in 

§§ 53.210, 53.220, 53.450(e), and 53.470 

since the last FSAR update; and 

(5) Analyses of new safety issues 

performed by or on behalf of the 

licensee at Commission request. 

(b)(1) The licensee must submit 

revisions containing updated 

information to the Commission, under 

§ 53.040, identifying the location of 

revised or new information. 

(2) The submittal must include— 

(i) A certification by a duly authorized 

officer of the licensee that either the 

information accurately presents changes 

made since the previous submittal, 

necessary to reflect information and 

analyses submitted to the Commission 

or prepared pursuant to Commission 

requirement, or that no such changes 

were made; and 

(ii) An identification of changes made 

under the provisions of § 53.1550 but 

not previously submitted to the 

Commission. 

(c) Each applicant for or holder of a 

COL under this part for which the 

Commission has not made the finding 

under § 53.1452(g) must submit an 

update to the FSAR annually by 

providing the information required in 

(a)(1) through (a)(5) of this section and 

meeting the requirements of paragraph 

(b) of this section. Combined license 

applicants who have requested the NRC 

to suspend its review of the COL 

application and COL holders who have 

informed the NRC that they do not plan 

to pursue construction need not submit 

an annual update of the FSAR. If a COL 

applicant requests that the NRC resume 

its review, or a COL holder notifies the 

NRC that the COL holder plans to 

commence or resume construction, then 

the COL applicant or holder must 

submit to NRC an update to its FSAR 

within 90 days of the request or 

notification, as applicable, and annually 

thereafter. 

(d) The FSAR (as updated) must be 

retained by the licensee until the 

Commission terminates its license. 

(e) Each holder of an ML under this 

part must submit an update of the FSAR 

reflecting any modification to the design 

that is directed or approved by the 

Commission under § 53.1288 or 

§ 53.1530, and any new analyses of the 

design requested by the Commission 

under § 53.1580. 

§ 53.1550 Evaluating changes to facility as 

described in Final Safety Analysis Reports. 

(a) The holder of an OL or COL may 

make changes in the facility as 

described in the FSAR (as updated) and 

make changes in the procedures as 

described in the FSAR (as updated) 

without obtaining a license amendment 

pursuant to § 53.1510 only if— 

(1) A change to the technical 

specifications incorporated in the 

license is not required; and 

(2) The change meets all of the 

following criteria: 

(i) Does not result in an increase to 

the frequency or consequences of an 

event sequence such that an event 

sequence not previously identified as 

risk significant becomes risk significant 

by the analyses performed in 

accordance with § 53.450(e). 

(ii) Does not result in an increase to 

the frequency or consequences of an 

event sequence such that an event 

sequence identified as risk significant in 

accordance with § 53.450(e) exceeds the 

licensing-basis event evaluation criteria 

required to be established in accordance 

with § 53.450(e). 

(iii) Does not involve either of the 

following: (A) a change to the NRCapproved comprehensive risk metric(s) 

or associated risk performance objective 

under § 53.220(b), or (B) an increase to 

the frequency or consequences of one or 

more event sequences such that there is 

more than a minimal reduction in the 

margin between the calculated 

comprehensive risks posed by the 

commercial nuclear plant and the safety 

criteria of § 53.220. 

(iv) Does not involve a departure from 

a method of evaluation described in the 

FSAR (as updated) used in assessing 

licensing-basis events in accordance 

with § 53.450 unless the results of the 

analysis under § 53.450 are conservative 

or essentially the same, the revised 

method of evaluation has been 

previously approved by the NRC for the 

intended application, or the revised 

method of evaluation can be used under 

an NRC-endorsed consensus code or 

standard. 

(v) Does not result in the escalation in 

the safety classification of an SSC from 

non-safety-related to non-safety-related 

but safety-significant or from non-safetyrelated but safety-significant to safetyrelated. 

(vi) Does not result in more than a 

minimal decrease in defense in depth. 

(vii) For commercial nuclear plants 

licensed under this part for which 

alternative evaluation criteria are 

adopted in accordance with § 53.470, 

does not result in a change to the 

frequency or consequences of event 

sequences such that the calculated 

margins between the results for event 

sequences evaluated in accordance with 

§ 53.450(e) and the alternative 

evaluation criteria decreases by 25 

percent or more. 

(viii) Does not result in the 

identification of a new design-basis 

accident in accordance with § 53.450(f). 

(ix) Does not result in a decrease by 

10 percent or more in the margin 

between the consequence of any designbasis accident and the safety criteria in 

§ 53.210. 

(x) Does not prevent meeting the 

design requirements in § 53.440(j) to 

limit the release of radionuclides from 

reactor systems, waste stores, or other 

significant inventories of radioactive 

materials assuming the impact of a 

large, commercial aircraft. 

(3) In implementing this paragraph, 

the FSAR (as updated) is considered to 

include FSAR changes since submittal 

of the last update of the FSAR under 

§ 53.1545. 

(4) The provisions in this section do 

not apply to changes to the facility or 

procedures when the applicable 

regulations establish more specific 

criteria for accomplishing such changes. 

(b)(1) A licensee who references a 

design certification rule may make 

departures from the standard design, 

without prior Commission approval, 

unless the proposed departure involves 

a change to the design as described in 

the rule certifying the design, in which 

case the requirements of § 53.1525 are 

applicable. 

(2) The licensee must maintain 

records of all departures from the 

certified design of the facility and these 

records must be maintained and 

available for audit until the termination 

of the license. The licensee must 

identify the location and nature of 

departures from licensing-basis 

information within supporting 

documents for a certified design within 

the updates to the Safety Analysis 

Report required by § 53.1545. 


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(3) Licensees for which the NRC has 

docketed the certifications required 

under § 53.1070 need not retain records 

of departures from the design of the 

facility associated with SSCs that have 

been permanently removed from service 

using an NRC-approved change process. 

(c)(1) The licensee must maintain 

records of changes in the facility and 

procedures made under paragraph (a) of 

this section. These records must include 

a written evaluation which provides the 

bases for the determination that the 

change does not require a license 

amendment under paragraph (a)(2) of 

this section. 

(2) The licensee must submit, as 

specified in § 53.040, a report 

containing a brief description of any 

departures and changes, including a 

summary of the evaluation of each. A 

report must be submitted at intervals 

not to exceed 24 months. For COLs, the 

report must be submitted at intervals 

not to exceed 6 months during the 

period from the date of application for 

a COL to the date the Commission 

makes its findings under § 53.1452(g). 

(3) The records of changes in the 

facility must be maintained until the 

termination of an OL or COL issued 

under this part, or the termination of a 

renewed license issued under 

§ 53.1595—whichever is later. Records 

of changes in procedures must be 

maintained for a period of 5 years. 

§ 53.1560 Updating program documents 

included in licensing-basis information. 

(a) Each holder under this part of an 

OL or COL for which the Commission 

has made the finding under § 53.1452(g) 

must biennially or more frequently 

update the program documents 

submitted as part of an application to 

obtain or maintain the license to assure 

that the information included in the 

documents contains the latest 

information developed. The submittals 

must include the effects on the content 

of the program documents of— 

(1) Changes made in the facility, 

procedures, licensee’s organization, or 

site environs; 

(2) Safety analyses and evaluations 

performed by the applicant or licensee 

either in support of approved license 

amendments or in support of 

conclusions that changes did not require 

a license amendment in accordance 

with § 53.1550; 

(3) Analyses of new safety issues 

performed by or on behalf of the 

licensee at Commission request; and 

(4) Changes to the programs as a result 

of operating experience, corrective 

actions, or other reasons deemed 

appropriate to ensure the programs 

serve their underlying purpose to 

support the requirements in subpart B of 

this part or other NRC regulations. 

(b)(1) The licensee must submit 

revisions containing updated 

information to the Commission, as 

specified in § 53.040, identifying the 

location of revised or new information. 

(2) The submittal must include— 

(i) A certification by a duly authorized 

officer of the licensee that either the 

information accurately presents changes 

made since the previous submittals, 

necessary to reflect information and 

analyses submitted to the Commission 

or prepared pursuant to Commission 

requirement, or that no such changes 

were made; and 

(ii) An identification of changes made 

under the provisions of § 53.1550 but 

not previously submitted to the 

Commission. 

(c) The updated program documents 

must be retained by the licensee until 

the Commission terminates their 

license. 

§ 53.1565 Evaluating changes to programs 

included in licensing-basis information. 

(a) A licensee may make changes to 

the facility, procedures, or organizations 

or address changes to site environs as 

described in the program documents 

included in licensing-basis information 

without obtaining prior NRC approval 

only if— 

(1) A change to the technical 

specifications incorporated in the 

license is not required; 

(2) An exemption from an NRC 

regulation is not required; and 

(3) The change conforms to programspecific requirements included in 

regulations in this part, technical 

specifications, or the NRC-approved 

program document included and 

reviewed as part of a license application 

under subpart H or an amendment 

under this subpart. 

(b) In implementing this section, the 

program documents (as updated) 

include changes since submittal of the 

last updates of the program documents 

pursuant to § 53.1560. 

(c) The provisions in this section do 

not apply to changes to the program 

documents when the applicable 

regulations establish more specific 

criteria for accomplishing such changes. 

(d) To make changes to the facility, 

procedures, or organizations or to 

address changes to site environs as 

described in the program documents 

included in licensing-basis information 

for individual programs, the following 

requirements must be satisfied: 

(1) Quality assurance program— 

operation. (i) Each holder under this 

part of an OL or COL, after the 

Commission makes the finding under 

§ 53.1452(g), may make a change to a 

previously accepted quality assurance 

program (QAP) description included or 

referenced in the Safety Analysis Report 

without prior NRC approval, provided 

the change does not reduce the 

commitments in the program 

description as accepted by the NRC. 

Changes to the QAP description that do 

not reduce the commitments must be 

submitted to the NRC in accordance 

with the requirements of § 53.1545. In 

addition to QAP changes involving 

administrative improvements and 

clarifications, spelling corrections, 

punctuation, or editorial items, the 

following changes are not considered to 

be reductions in commitment: 

(A) The use of a quality assurance 

(QA) standard approved by the NRC 

which is more recent than the QA 

standard in the licensee’s QAP at the 

time of the change; 

(B) The use of a QA alternative or 

exception approved by an NRC safety 

evaluation, provided that the bases of 

the NRC approval are applicable to the 

licensee’s facility; 

(C) The use of generic organizational 

position titles that clearly denote the 

position function, supplemented as 

necessary by descriptive text, rather 

than specific titles; 

(D) The use of generic organizational 

charts to indicate functional 

relationships, authorities, and 

responsibilities, or, alternately, the use 

of descriptive text; 

(E) The elimination of QAP 

information that duplicates language in 

QA regulatory guides and QA standards 

to which the licensee is committed; and 

(F) Organizational revisions that 

ensure that persons and organizations 

performing QA functions continue to 

have the requisite authority and 

organizational freedom, including 

sufficient independence from cost and 

schedule when opposed to safety 

considerations. 

(ii) Changes to the QAP description 

that do reduce the commitments must 

be submitted to the NRC and receive 

NRC approval prior to implementation, 

as follows: 

(A) Changes made to the QAP 

description as presented in the Safety 

Analysis Report or in a topical report 

must be submitted as specified in 

§ 53.040. 

(B) The submittal of a change to the 

Safety Analysis Report QAP description 

must include all pages affected by that 

change and must be accompanied by a 

forwarding letter identifying the change, 

the reason for the change, and the basis 

for concluding that the revised program 

incorporating the change continues to 

satisfy the criteria of appendix B to part 


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50 of this chapter and the Safety 

Analysis Report QAP description 

commitments previously accepted by 

the NRC (the letter need not provide the 

basis for changes that correct spelling, 

punctuation, or editorial items). 

(C) A copy of the forwarding letter 

identifying the change must be 

maintained as a facility record for 3 

years. 

(D) Changes to the QAP description 

included or referenced in the Safety 

Analysis Report shall be regarded as 

accepted by the Commission upon 

receipt of a letter to this effect from the 

appropriate reviewing office of the 

Commission or 60 days after submittal 

to the Commission, whichever occurs 

first. 

(2) Quality assurance program— 

siting, construction, and manufacturing. 

Each holder of an LWA, early site 

permit, CP, ML, or COL, before the 

Commission makes the finding under 

§ 53.1452(g) of this chapter, under this 

part may make a change to a previously 

accepted QAP description included or 

referenced in the Safety Analysis Report 

without prior NRC approval, provided 

the change does not reduce the 

commitments in the program 

description previously accepted by the 

NRC. Changes to the QAP description 

that do not reduce the commitments 

must be submitted to NRC within 90 

days. Changes to the QAP description 

that reduce the commitments must be 

submitted to NRC and receive NRC 

approval before implementation, as 

follows: 

(i) Changes to the Safety Analysis 

Report must be submitted for review as 

specified in § 53.040. Changes made to 

NRC-accepted QA topical report 

descriptions must be submitted as 

specified in § 53.040. 

(ii) The submittal of a change to the 

Safety Analysis Report QAP description 

must include all pages affected by that 

change and must be accompanied by a 

forwarding letter identifying the change, 

the reason for the change, and the basis 

for concluding that the revised program 

incorporating the change continues to 

satisfy the criteria of appendix B of part 

50 of this chapter and the Safety 

Analysis Report QAP description 

commitments previously accepted by 

the NRC (the letter need not provide the 

basis for changes that correct spelling, 

punctuation, or editorial items). 

(iii) A copy of the forwarding letter 

identifying the changes must be 

maintained as a facility record for 3 

years. 

(iv) Changes to the QAP description 

included or referenced in the Safety 

Analysis Report shall be regarded as 

accepted by the Commission upon 

receipt of a letter to this effect from the 

appropriate reviewing office of the 

Commission or 60 days after submittal 

to the Commission, whichever occurs 

first. 

(3) Emergency preparedness program. 

(i) Definitions for the purpose of 

paragraph (d)(3) of this section: 

(A) Change means an action that 

results in modification or addition to, or 

removal from, the licensee’s emergency 

plan. All such changes are subject to the 

provisions of this section except where 

the applicable regulations establish 

specific criteria for accomplishing a 

particular change. 

(B) Emergency plan means the 

document(s), prepared and maintained 

by the licensee, that identify and 

describe the licensee’s methods for 

maintaining emergency preparedness 

and responding to emergencies. An 

emergency plan includes the plan as 

originally approved by the NRC and all 

subsequent changes made by the 

licensee with, and without, prior NRC 

review and approval under paragraph 

(d)(3) of this section. 

(C) Emergency planning function 

means a capability or resource necessary 

to prepare for and respond to a 

radiological emergency. 

(D) Reduction in effectiveness means 

a change in an emergency plan that 

results in reducing the licensee’s 

capability to perform an emergency 

planning function in the event of a 

radiological emergency. 

(ii)(A) Except as provided in 

paragraph (d)(3)(ii)(B) of this section, a 

holder of an OL under this part, or a 

COL under this part after the 

Commission makes the finding under 

§ 53.1452(g), must follow and maintain 

the effectiveness of an emergency plan 

that meets the requirements in appendix 

E to part 50 of this chapter and the 

planning standards of § 50.47(b). 

(B) A holder of an OL under this part 

for a commercial nuclear plant 

consisting of small modular reactors 

(SMRs) or non-light-water reactors, or a 

holder of a COL under this part after the 

Commission makes the finding under 

§ 53.1452(g) for a commercial nuclear 

plant consisting of either SMRs or nonlight-water reactors, must follow and 

maintain the effectiveness of either an 

emergency plan that meets the 

requirements in § 50.160 or an 

emergency plan that meets the 

requirements in appendix E to part 50 

of this chapter and the planning 

standards of § 50.47(b). 

(iii)(A) Except as provided in 

paragraph (d)(3)(iii)(B) of this section, 

the licensee may make changes to its 

emergency plan without NRC approval 

only if the licensee performs and retains 

an analysis demonstrating that the 

changes do not reduce the effectiveness 

of the plan and the plan, as changed, 

continues to meet the requirements in 

appendix E to part 50 of this chapter 

and the planning standards of 

§ 50.47(b). 

(B) A license under this part for a 

commercial nuclear plant consisting of 

either SMRs or non-light-water reactors 

may make changes to its emergency 

plan without NRC approval only if the 

licensee performs and retains an 

analysis demonstrating that the changes 

do not reduce the effectiveness of the 

plan and the plan, as changed, 

continues to meet either the 

requirements in § 50.160 or the 

requirements in appendix E to part 50 

and the planning standards of 

§ 50.47(b). 

(iv) The changes to a licensee’s 

emergency plan that reduce the 

effectiveness of the plan as defined in 

paragraph (d)(3)(i)(D) of this section 

may not be implemented without prior 

approval by the NRC. A licensee 

desiring to make such a change must 

submit an application for an 

amendment to its license. In addition to 

the filing requirements of §§ 53.1510 

and 53.1515, the request must include 

all emergency plan pages affected by 

that change and must be accompanied 

by a forwarding letter identifying the 

change, the reason for the change, and 

the basis for concluding that the 

licensee’s emergency plan, as revised, 

will continue to meet either the 

requirements in § 50.160 to this chapter 

or the requirements in appendix E to 

part 50 of this chapter and the planning 

standards of § 50.47(b) of this chapter. 

(v) The licensee must retain a record 

of each change to the emergency plan 

made without prior NRC approval for a 

period of three years from the date of 

the change and shall submit, as 

specified in § 53.040, a report of each 

such change, including a summary of its 

analysis, within 30 days after the change 

is put in effect. 

(vi) The licensee must retain the 

emergency plan and each change for 

which prior NRC approval was obtained 

pursuant to paragraph (d)(3)(iv) of this 

section as a record until the 

Commission terminates the license for 

the nuclear power reactor. 

(vii)(A) The licensee must provide for 

the development, revision, 

implementation, and maintenance of its 

emergency preparedness program. The 

licensee must ensure that all program 

elements are reviewed by persons who 

have no direct responsibility for the 

implementation of the emergency 

preparedness program either— 


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(1) At intervals not to exceed 12 

months; or 

(2) As necessary, based on an 

assessment by the licensee against 

performance indicators, and as soon as 

reasonably practicable after a change 

occurs in personnel, procedures, 

equipment, or facilities that potentially 

could adversely affect emergency 

preparedness, but no longer than 12 

months after the change. In any case, all 

elements of the emergency preparedness 

program must be reviewed at least once 

every 24 months. 

(B) The review must include an 

evaluation for adequacy of interfaces 

with State participating Tribal and local 

governments and of licensee drills, 

exercises, capabilities, and procedures. 

The results of the review, along with 

recommendations for improvements, 

must be documented, reported to the 

licensee’s corporate and plant 

management, and retained for a period 

of 5 years. The part of the review 

involving the evaluation for adequacy of 

interface with State, participating Tribal 

and local governments must be available 

to the appropriate State, participating 

Tribal and local governments. 

(4) Security programs. (i) The licensee 

must prepare and maintain safeguards 

contingency plan procedures in 

accordance with appendix C of part 73 

of this chapter for affecting the actions 

and decisions contained in the 

Responsibility Matrix of the safeguards 

contingency plan. The licensee may not 

make a change that would decrease the 

safeguard effectiveness of a physical 

security plan, or guard training and 

qualification plan, or cybersecurity plan 

submitted under subpart H or part 73 of 

this chapter, or of the first four 

categories of information (Background, 

Generic Planning Base, Licensee 

Planning Base, Responsibility Matrix) 

contained in a licensee safeguards 

contingency plan submitted under 

subpart H or part 73 of this chapter, as 

applicable, without prior approval of 

the Commission. A licensee desiring to 

make such a change must submit an 

application for amendment to the 

licensee’s license under §§ 53.1510, 

53.1515, and 53.1520. 

(ii) The licensee may make changes to 

the plans referenced in paragraph (4)(i) 

of this section without prior 

Commission approval if the changes do 

not decrease the safeguards 

effectiveness of the plan. The licensee 

must maintain records of changes to the 

plans made without prior Commission 

approval for a period of 3 years from the 

date of the change, and must submit, as 

specified in § 53.040, a report 

containing a description of each change 

within 2 months after the change is 

made. Prior to the safeguards 

contingency plan being put into effect, 

the licensee must have— 

(A) All safeguards capabilities 

specified in the safeguards contingency 

plan available and functional; 

(B) Detailed procedures developed 

according to appendix C to part 73 of 

this chapter available at the licensee’s 

site; and 

(C) All appropriate personnel trained 

to respond to safeguards incidents as 

outlined in the plan and specified in the 

detailed procedures. 

(iii) The licensee must provide for the 

development, revision, implementation, 

and maintenance of its safeguards 

contingency plan. The licensee must 

ensure that all program elements are 

reviewed by individuals independent of 

both security program management and 

personnel who have direct 

responsibility for implementation of the 

security program either— 

(A) At intervals not to exceed 12 

months; or 

(B) As necessary, based on an 

assessment by the licensee against 

performance indicators, and as soon as 

reasonably practicable after a change 

occurs in personnel, procedures, 

equipment, or facilities that potentially 

could adversely affect security, but no 

longer than 12 months after the change. 

In any case, all elements of the 

safeguards contingency plan must be 

reviewed at least once every 24 months. 

(iv) The review must include a review 

and audit of safeguards contingency 

procedures and practices, an audit of 

the security system testing and 

maintenance program, and a test of the 

safeguards systems along with 

commitments established for response 

by local law enforcement authorities. 

The results of the review and audit, 

along with recommendations for 

improvements, must be documented, 

reported to the licensee’s corporate and 

plant management, and kept available at 

the plant for inspection for a period of 

3 years. 

§ 53.1570 Transfer of licenses. 

(a) No commercial nuclear plant 

license issued under this part, or any 

right thereunder, shall be transferred, 

assigned, or in any manner disposed of, 

either voluntarily or involuntarily, 

directly or indirectly, through transfer of 

control of the license to any person, 

unless the Commission gives its consent 

in writing. 

(b)(1) An application for transfer of a 

license must include— 

(i) As much of the information 

described in §§ 53.1109, 53.1306, 

53.1366, and 53.1413 with respect to the 

identity and technical and financial 

qualifications of the proposed transferee 

as would be required by those sections 

if the application were for an initial 

license. The Commission may require 

additional information such as data 

respecting proposed safeguards against 

hazards from radioactive materials and 

the applicant’s qualifications to protect 

against such hazards. 

(ii) A statement of the purposes for 

which the transfer of the license is 

requested, the nature of the transaction 

necessitating or making desirable the 

transfer of the license, and an agreement 

to limit access to Restricted Data or 

Classified National Security Information 

pursuant to § 53.1115. The Commission 

may require any person who submits an 

application for license pursuant to the 

provisions of this section to file a 

written consent from the existing 

licensee or a certified copy of an order 

or judgment of a court of competent 

jurisdiction attesting to the person’s 

right (subject to the licensing 

requirements of the Act and these 

regulations) to possession of the facility 

or site involved. 

(2) [Reserved] 

(c) After appropriate notice to 

interested persons, including the 

existing licensee, and observance of 

such procedures as may be required by 

the Act or regulations or orders of the 

Commission, the Commission will 

approve an application for the transfer 

of a license, if the Commission 

determines— 

(1) That the proposed transferee is 

qualified to be the holder of the license; 

and 

(2) That transfer of the license is 

otherwise consistent with applicable 

provisions of law, regulations, and 

orders issued by the Commission 

pursuant thereto. 

§ 53.1575 Termination of licenses. 

(a) When the holder of an OL or COL 

under this part has determined to 

permanently cease operations the 

licensee must, within 30 days, submit a 

written certification to the NRC, 

consistent with the requirements of 

§ 53.1070. 

(b) Once fuel has been permanently 

removed from the reactor system, the 

licensee must submit a written 

certification to the NRC that meets the 

requirements of § 53.1070. 

(c)(1) Upon docketing of the 

certifications for permanent cessation of 

operations and permanent removal of 

fuel from the reactor system, or when a 

final legally effective order to 

permanently cease operations has come 

into effect, the license no longer 

authorizes operation of the reactor or 


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emplacement or retention of fuel into 

the reactor system. 

(2) Activities associated with 

decommissioning will be carried out in 

accordance with the requirements and 

procedures in subpart G of this part. 

(3) The Commission shall terminate 

the license if it determines that— 

(i) The remaining dismantlement has 

been performed in accordance with the 

approved license termination plan 

required in subpart G of this part; and 

(ii) The final radiation survey and 

associated documentation, including an 

assessment of dose contributions 

associated with parts released for use 

before approval of the license 

termination plan, demonstrate that the 

facility and site have met the criteria for 

decommissioning in subpart E of 10 

CFR part 20. 

(d) A holder of a CP or COL under this 

part may request the termination of the 

license as well as licenses issued by the 

NRC under parts 30, 40, or 70 of this 

chapter prior to plant operations. Such 

requests may support an immediate 

NRC approval of the site for unrestricted 

use. 

§ 53.1580 Information requests. 

Each licensee under this part must at 

any time before termination of the 

license, upon request of the 

Commission, submit, as specified in 

§ 53.040 written statements, signed 

under oath or affirmation, to enable the 

Commission to determine whether or 

not the license should be modified, 

suspended, or revoked. Except for 

information sought to verify licensee 

compliance with the current licensing 

basis for that facility, the NRC must 

prepare the reason or reasons for each 

information request prior to issuance to 

ensure that the burden to be imposed on 

respondents is justified in view of the 

potential safety significance of the issue 

to be addressed in the requested 

information. Each such justification 

provided for an evaluation performed by 

the NRC staff must be approved by the 

Executive Director for Operations or his 

or her designee prior to issuance of the 

request. 

§ 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 § 53.1342(b); or for 

violation of, or failure to observe, any of 

the terms and provisions of the Act, 

regulations, license, approval, or order 

of the Commission. 

§ 53.1590 Backfitting. 

(a)(1) Backfitting means the 

modification of or addition to systems, 

structures, components, or design of a 

facility; or the design approval or ML for 

a facility; or the procedures or 

organization required to design, 

construct or operate a facility; any of 

which may result from a new or 

amended provision in the Commission’s 

regulations or the imposition of a 

regulatory staff position interpreting the 

Commission’s regulations that is either 

new or different from a previously 

applicable staff position after the date of 

the commercial nuclear plant license 

issued under this part. 

(2) Except as provided in paragraph 

(a)(4) of this section, the Commission 

shall require a systematic and 

documented analysis pursuant to 

paragraph (b) of this section for backfits 

which it seeks to impose. 

(3) Except as provided in paragraph 

(a)(4) of this section, the Commission 

shall require the backfitting of a facility 

only when it determines, based on the 

analysis described in paragraph (b) of 

this section, that there is a substantial 

increase in the overall protection of the 

public health and safety or the common 

defense and security to be derived from 

the backfit and that the direct and 

indirect costs of implementation for that 

facility are justified in view of this 

increased protection. 

(4) The provisions of paragraphs (a)(2) 

and (a)(3) of this section are 

inapplicable and, therefore, backfit 

analysis is not required and the 

standards in paragraph (a)(3) of this 

section do not apply where the 

Commission or staff, as appropriate, 

finds and declares, with appropriate 

documented evaluation for its finding, 

either— 

(i) That a modification is necessary to 

bring a facility into compliance with a 

license or the rules or orders of the 

Commission, or into conformance with 

written commitments by the licensee; or 

(ii) That regulatory action is necessary 

to ensure that the facility provides 

adequate protection to the health and 

safety of the public and is in accord 

with the common defense and security; 

or 

(iii) That the regulatory action 

involves defining or redefining what 

level of protection to the public health 

and safety or common defense and 

security should be regarded as adequate. 

(5) The Commission must always 

require the backfitting of a facility if it 

determines that such regulatory action 

is necessary to ensure that the facility 

provides adequate protection to the 

health and safety of the public and is in 

accord with the common defense and 

security. 

(6) The documented evaluation 

required by paragraph (a)(4) of this 

section must include a statement of the 

objectives of and reasons for the 

modification and the basis for invoking 

the exception. If immediately effective 

regulatory action is required, then the 

documented evaluation may follow 

rather than precede the regulatory 

action. 

(7) If there are two or more ways to 

achieve compliance with a license or 

the rules or orders of the Commission, 

or with written licensee commitments, 

or there are two or more ways to reach 

a level of protection which is adequate, 

then ordinarily the applicant or licensee 

is free to choose the way which best 

suits its purposes. However, should it be 

necessary or appropriate for the 

Commission to prescribe a specific way 

to comply with its requirements or to 

achieve adequate protection, then cost 

may be a factor in selecting the way, 

provided that the objective of 

compliance or adequate protection is 

met. 

(b) In reaching the determination 

required by paragraph (a)(3) of this 

section, the Commission will consider 

how the backfit should be scheduled in 

light of other ongoing regulatory 

activities at the facility and, in addition, 

will consider information available 

concerning any of the following factors 

as may be appropriate and any other 

information relevant and material to the 

proposed backfit: 

(1) The statement of the specific 

objectives that the proposed backfit is 

designed to achieve; 

(2) The general description of the 

activity that would be required by the 

licensee or applicant in order to 

complete the backfit; 

(3) The potential change in the risk to 

the public from the accidental off-site 

release of radioactive material; 

(4) The potential impact on 

radiological exposure of facility 

employees; 


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(5) The installation and continuing 

costs associated with the backfit, 

including the cost of facility downtime 

or the cost of construction delay; 

(6) The potential safety impact of 

changes in plant or operational 

complexity, including the relationship 

to proposed and existing regulatory 

requirements; 

(7) The estimated resource burden on 

the NRC associated with the proposed 

backfit and the availability of such 

resources; 

(8) The potential impact of differences 

in facility type, design or age on the 

relevancy and practicality of the 

proposed backfit; 

(9) Whether the proposed backfit is 

interim or final and, if interim, the 

justification for imposing the proposed 

backfit on an interim basis. 

(c) No licensing action will be 

withheld during the pendency of backfit 

analyses required by the Commission’s 

rules. 

(d) The Executive Director for 

Operations shall be responsible for 

implementation of this section, and all 

analyses required by this section shall 

be approved by the Executive Director 

for Operations or his or her designee. 

§ 53.1595 Renewal. 

Licenses may be renewed by the 

Commission upon expiration of the 

period of the license. 

Subpart J—Reporting and Other 

Administrative Requirements 

§ 53.1600 General information. 

Each applicant and licensee under 

this part must ensure that U.S. Nuclear 

Regulatory Commission (NRC) 

inspectors have unfettered access to 

sites and facilities licensed or proposed 

to be licensed in § 53.1610, must 

maintain records and make reports to 

the NRC in accordance with 

requirements in §§ 53.1620 through 

53.1650, must satisfy financial 

qualification and reporting requirements 

in §§ 53.1660 through 53.1700, and 

must obtain and maintain required 

financial protections in case of an 

accident in §§ 53.1720 and 53.1730. 

§ 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. 

§ 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 § 53.040. 

(b) [Reserved] 

(c) Records that are required by the 

regulations in this part, by license 

condition, or by technical specifications 

must be retained for the period specified 

by the appropriate regulation, license 

condition, or technical specification. If 

a retention period is not otherwise 

specified, these records must be 

retained until the Commission 

terminates the facility license or, in the 

case of an early site permit, until the 

permit expires. 

(d)(1) Records which must be retained 

under this part may be the original or a 

reproduced copy or a microform if the 

reproduced copy or microform is duly 

authenticated by authorized personnel 

and the microform is capable of 

producing a clear and legible copy after 

storage for the period specified by 

Commission regulations. The record 

may also be stored in electronic media 

with the capability of producing legible, 

accurate, and complete records during 

the required retention period. Records 

such as letters, drawings, and 

specifications, must include all 

pertinent information such as stamps, 

initials, and signatures. The licensee 

must maintain adequate safeguards 

against tampering with, and loss of 

records. 

(2) If there is a conflict between the 

Commission’s regulations in this part, 

license condition, or technical 

specification, or other written 

Commission approval or authorization 

pertaining to the retention period for the 

same type of record, the retention 

period specified in the regulations in 

this part for such records shall apply 

unless the Commission, under § 53.080 

of this part, has granted a specific 

exemption from the record retention 

requirements in the regulations in this 

part. 

(e) Each licensee must notify the 

Commission as specified in § 53.040 of 

this part, of successfully completing 

power ascension testing or startup 

testing as applicable within 30 calendar 

days of completing the testing. 

§ 53.1630 Immediate notification 

requirements for operating commercial 

nuclear plants. 

(a) General requirements.1 (1) Each 

holder of an OL under this part or a COL 

under this part after the Commission 

makes the finding under § 53.1452(g), 

must notify the NRC Operations Center 


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via the Emergency Notification System 

(ENS) of— 

(i) The declaration of any of the 

Emergency Classes specified in the 

licensee’s approved Emergency Plan; or 

(ii) Those non-emergency events 

specified in paragraph (b) of this section 

that occurred within 3 years of the date 

of discovery. 

(2) If the ENS is inoperative, the 

licensee must make the required 

notifications via commercial telephone 

service, other dedicated telephone 

system, or any other method which will 

ensure that a report is made as soon as 

practical to the NRC Headquarters 

Operations Center at the numbers 

specified in appendix A to part 73 of 

this chapter. 

(3) The licensee must notify the NRC 

immediately after notification of the 

appropriate State or local agencies and 

not later than 1 hour after the time the 

licensee declares one of the Emergency 

Classes. 

(4) The licensee must activate the data 

links with the NRC as specified in their 

emergency plans after declaring an 

Emergency Class for events of actual or 

potential substantial degradation of 

plant safety or security, probable risk to 

site personnel life, or site equipment 

damage caused by hostile action. The 

data links may also be activated by the 

licensee during emergency drills or 

exercises if the licensee’s computer 

system has the capability to transmit the 

exercise data. 

(5) When making a report under 

paragraph (a)(1) of this section, the 

licensee must identify— 

(i) The Emergency Class declared; or 

(ii) Paragraph (b)(1), ‘‘One-hour 

reports,’’ paragraph (b)(2), ‘‘Four-hour 

reports,’’ or paragraph (b)(3), ‘‘Eighthour reports,’’ as the paragraph of this 

section requiring notification of the nonemergency event. 

(b) Non-emergency events. (1) Onehour reports. If not reported as a 

declaration of an Emergency Class 

under paragraph (a) of this section, the 

licensee must notify the NRC as soon as 

practical and in all cases within one 

hour of the occurrence of any deviation 

from the plant’s Technical 

Specifications authorized under 

§ 53.740(h) of this part. 

(2) Four-hour reports. If not reported 

under paragraphs (a) or (b)(1) of this 

section, the licensee must notify the 

NRC as soon as practical, and in all 

cases, within 4 hours of the occurrence 

of any of the following: 

(i) The initiation of any commercial 

nuclear plant shutdown required by the 

plant’s Technical Specifications. 

(ii) Any event or condition that results 

in actuation of the reactor protection 

system when the reactor is critical 

except when the actuation results from 

and is part of a pre-planned sequence 

during testing or reactor operation. 

(iii) Any event or condition that 

results in an unplanned actuation of a 

safety-related (SR) standby cooling 

system or the unplanned sole reliance 

on an SR standby cooling system for 

those systems that are in constant 

operation. 

(iv) Any event or condition that 

results in an unplanned movement of, 

change of state in, or chemical 

interaction involving a significant 

amount of radioactive material within 

the commercial nuclear plant. 

(v) Any event or situation, related to 

the health and safety of the public or 

onsite personnel, or protection of the 

environment, for which a news release 

is planned or notification to other 

government agencies has been or will be 

made. Such an event may include an 

onsite fatality or inadvertent release of 

radioactively contaminated materials. 

(3) Eight-hour reports. If not reported 

under paragraphs (a), (b)(1), or (b)(2) of 

this section, the licensee must notify the 

NRC as soon as practical and in all cases 

within 8 hours of the occurrence of any 

of the following: 

(i) Any event or condition that results 

in— 

(A) The condition of the commercial 

nuclear plant, including its principal 

safety barriers, being seriously 

degraded; or 

(B) The commercial nuclear plant 

being in a condition not analyzed under 

§ 53.450 that significantly degrades 

plant safety. 

(ii) Any event or condition that results 

in valid actuation of an SR system, 

except when the actuation results from 

and is part of a pre-planned sequence 

during testing or reactor operation. 

(iii) Any event or condition that at the 

time of discovery could have prevented 

the fulfilment of the safety functions 

identified under § 53.230. Events 

covered may include one or more 

procedural errors, equipment failures, 

and/or discovery of design, analysis, 

fabrication, construction, and/or 

procedural inadequacies. However, 

individual component failures need not 

be reported pursuant to this paragraph 

if other equipment was operable and 

available to perform the required safety 

function. 

(iv) Any event requiring the transport 

of a radioactively contaminated person 

to an offsite medical facility for 

treatment. 

(v) Any event that results in a major 

loss of emergency assessment capability, 

offsite response capability, or offsite 

communications capability (e.g., 

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. 

1Other requirements for immediate 

notification of the NRC by licensed operating 

commercial nuclear plants are contained 

elsewhere in this chapter, in particular 

§§ 20.1906, 20.2202, 72.216, 73.77, and 

73.1200 of this chapter. 

§ 53.1640 Licensee event report system. 

(a) Reportable events. (1) Each 

commercial nuclear plant licensee 

holding an OL under this part or a COL 

under this part after the Commission 

makes the finding under § 53.1452(g), 

must submit a Licensee Event Report 

(LER) for any event of the type 

described in this paragraph within 60 

days after discovery of the event. In the 

case of an invalid actuation reported 

under § 53.1640(a)(2), other than 

automatic reactor shutdown when the 

reactor is critical, the licensee may, at 

its option, provide a telephone 

notification to the NRC Operations 

Center within 60 days after discovery of 

the event instead of submitting a written 

LER. Unless otherwise specified in this 

section, the licensee must report an 

event if it occurred within 3 years of the 

date of discovery regardless of the plant 

mode or power level, and regardless of 

the significance of the structure, system, 

or component that initiated the event. 

(2) The licensee must report— 

(i)(A) The completion of any 

commercial nuclear plant shutdown 

required by the plant’s Technical 

Specifications. 


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(B) Any operation or condition which 

was prohibited by the plant’s Technical 

Specifications except when— 

(1) The Technical Specification is 

administrative in nature; 

(2) The event consisted solely of a 

case of a late surveillance test where the 

oversight was corrected, the test was 

performed, and the equipment was 

found to be capable of performing its 

specified safety functions; or 

(3) The Technical Specification was 

revised prior to discovery of the event 

such that the operation or condition was 

no longer prohibited at the time of the 

event. 

(C) Any deviation from the plant’s 

Technical Specifications authorized 

under § 53.740(h). 

(ii) Any event or condition that 

resulted in— 

(A) The condition of the commercial 

nuclear plant, including its principal 

safety barriers, being seriously 

degraded; or 

(B) The commercial nuclear plant 

being in a condition not analyzed under 

§ 53.450 that significantly degrades 

plant safety. 

(iii) Any natural phenomena or other 

external condition that posed an actual 

threat to the safety of the commercial 

nuclear plant or significantly hampered 

site personnel in the performance of 

duties necessary for the safe operation 

of the commercial nuclear plant. 

(iv) Any event or condition that 

resulted in inadvertent operation of any 

structures, systems, and component 

classified as SR for an identified safety 

function under § 53.460 or the 

unplanned sole reliance on an SR 

system for those systems that are in 

constant operation, except when— 

(A) The actuation resulted from and 

was part of a pre-planned sequence 

during testing; or 

(B) The actuation was invalid and— 

(1) Occurred while the system was 

properly removed from service; or 

(2) Occurred after the safety function 

had been already completed. 

(v) Any event or condition that could 

have prevented the fulfillment of the 

safety functions identified under 

§ 53.230. 

(vi) Events covered in paragraph 

(a)(2)(v) of this section may include one 

or more procedural errors, equipment 

failures, and/or discovery of design, 

fabrication, construction, and/or 

procedural inadequacies. However, 

individual component failures need not 

be reported pursuant to paragraph 

(a)(2)(v) of this section if any other 

equipment was operable and available 

to perform the required safety function. 

(vii)(A) Any event or condition that as 

a result of a single cause could have 

prevented the fulfillment of any of the 

safety functions identified under 

§ 53.230. 

(B) Events covered in paragraph 

(a)(2)(vii)(A) of this section may include 

cases of procedural error, equipment 

failure, and/or discovery of a design, 

analysis, fabrication, construction, 

and/or procedural inadequacy. 

However, licensees are not required to 

report an event pursuant to paragraph 

(a)(2)(vii)(A) of this section if the event 

results from— 

(1) A shared dependency among 

trains or channels that is a natural or 

expected consequence of the approved 

plant design; or 

(2) Normal and expected wear or 

degradation. 

(viii)(A) Any airborne radioactive 

release that, when averaged over a time 

period of 1-hour, resulted in airborne 

radionuclide concentrations in an 

unrestricted area that exceeds 20 times 

the applicable concentration limits 

specified in appendix B to 10 CFR part 

20, table 2, column 1. 

(B) Any liquid effluent release that, 

when averaged over a time period of 

1-hour, exceeds 20 times the applicable 

concentrations specified in appendix B 

to 10 CFR part 20, table 2, column 2, at 

the point of entry into the receiving 

waters (i.e., unrestricted area) for all 

radionuclides except tritium and 

dissolved noble gases. 

(ix) Any event that posed an actual 

threat to the safety of the commercial 

nuclear plant or significantly hampered 

site personnel in the performance of 

duties necessary for the safe operation 

of the plant, including fires, toxic gas 

releases, or radioactive releases. 

(b) Contents. The LER must contain— 

(1) A brief abstract describing the 

major occurrences during the event, 

including all component or system 

failures that contributed to the event 

and significant corrective action taken 

or planned to prevent recurrence. 

(2)(i) A clear, specific narrative 

description of what occurred so that 

knowledgeable readers conversant with 

the design of commercial nuclear plants, 

but not familiar with the details of a 

particular plant, can understand the 

complete event. 

(ii) The narrative description must 

include the following specific 

information as appropriate for the 

particular event: 

(A) Plant operating conditions before 

the event. 

(B) Status of systems, structures, or 

components that were inoperable at the 

start of the event and that contributed to 

the event. 

(C) Dates and approximate time of the 

occurrences. 

(D) The cause of each component or 

system failure or personnel error, if 

known. 

(E) The failure mode, mechanism, and 

effect of each failed component, if 

known. 

(F) [Reserved] 

(G) For failures of components with 

multiple functions, include a list of 

systems or secondary functions that 

were also affected. 

(H) For failure that rendered a 

component or system classified as SR or 

non-safety-related but safety-significant 

inoperable, an estimate of the elapsed 

time from the discovery of the failure 

until the component or system was 

returned to service. 

(I) The method of discovery of each 

component or system failure or 

procedural error. 

(J) For each human performance 

related root cause, the licensee must 

discuss the cause(s) and circumstances. 

(K) Automatically and manually 

initiated safety system responses. 

(L) The manufacturer and model 

number (or other identification) of each 

component that failed during the event. 

(3) An assessment of the safety 

consequences and implications of the 

event. This assessment must include— 

(i) The availability of systems or 

components that could have performed 

the same function as the components 

and systems that failed during the event, 

and 

(ii) For events that occurred when the 

reactor was shut down, the availability 

of systems or components that are 

needed to shut down the reactor and 

maintain safe shutdown conditions, 

remove residual heat, control the release 

of radioactive material, or mitigate the 

consequences of an accident. 

(4) A description of any corrective 

actions planned as a result of the event, 

including those to reduce the 

probability of similar events occurring 

in the future. 

(5) Reference to any previous similar 

events at the same plant that are known 

to the licensee. 

(6) The name and contact information 

of a person within the licensee’s 

organization who is knowledgeable 

about the event and can provide 

additional information concerning the 

event and the plant’s characteristics. 

(c) Supplemental Information. The 

Commission may require the licensee to 

submit specific additional information 

beyond that required by paragraph (b) of 

this section if the Commission finds that 

supplemental material is necessary for 

complete understanding of an unusually 

complex or significant event. These 

requests for supplemental information 

will be made in writing and the licensee 


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must submit, as specified in § 53.040, 

the requested information as a 

supplement to the initial LER. 

(d) Submission of Reports. Licensee 

Event Reports must be prepared on 

Form NRC 366 and submitted to the 

NRC, as specified in § 53.040. 

(e) Report Legibility. The reports and 

copies that licensees are required to 

submit to the Commission under the 

provisions of this section must be of 

sufficient quality to permit legible 

reproduction and micrographic 

processing. 

53.1645 Reports of radiation exposure to 

members of the public. 

(a) Each holder of an OL, and each 

holder of a COL after the Commission 

has made the finding under 

§ 53.1452(g), must submit radiological 

reports as required by 10 CFR part 20, 

as well as an Annual Radioactive 

Effluent Release Report and an Annual 

Radiological Environmental Operating 

Report. The Annual Radioactive 

Effluent Release Report must specify the 

quantity of each of the principal 

radionuclides released to unrestricted 

areas in liquid and in gaseous effluents 

and an estimate of the dose received by 

the maximally exposed member of the 

public in an unrestricted area from 

effluents and direct radiation from 

contained sources during the previous 

calendar year. The Annual Radiological 

Environmental Operating Report must 

provide data on measurable levels of 

radiation and radioactive materials in 

the environment, must include an 

evaluation of the relationship between 

quantities of radioactive material 

released in effluents and resultant 

radiation doses to individuals from 

principal pathways of exposure, and 

must include the results of 

environmental monitoring during the 

previous calendar year. These reports 

must also include any other information 

as may be required by the Commission 

to estimate maximum potential annual 

radiation doses to the public. The 

reports must be submitted as specified 

in § 53.040 by May 15 of each 

successive year. If the total effective 

dose equivalent to members of the 

public in unrestricted areas during the 

reporting period is greater than the as 

low as is reasonably achievable 

(ALARA) design objectives established 

under § 53.425, the report must specify 

the causes for exceeding the ALARA 

design objective and describe any 

corrective actions. On the basis of these 

reports and any additional information 

the Commission may obtain from the 

licensee or others, the Commission may 

require the licensee to take action as the 

Commission deems appropriate. 

(b) If during any calendar quarter the 

radiation exposure to a member of the 

public in the unrestricted areas, 

calculated on the same basis as the 

respective ALARA design objective 

exposure, exceeds one-half of the 

annual ALARA design objective 

exposure, the licensee must submit a 

report as specified in § 53.040. The 

report shall specify the causes for 

exceeding one-half the annual ALARA 

design objective exposure in a quarter 

and describe corrective actions that the 

licensee will take to maintain radiation 

exposure to levels within the ALARA 

design objectives for the remainder of 

the year. The report shall be submitted 

within 30 days from the end of the 

quarter when one-half of the annual 

ALARA design objective exposure was 

exceeded. 

§ 53.1650 Facility information and 

verification. 

(a) In response to a written request by 

the Commission, each applicant for a CP 

or license and each recipient of a CP or 

a license must submit facility 

information, as described in § 75.10 of 

this chapter, on International Atomic 

Energy Agency (IAEA) Design 

Information Questionnaire forms and 

site information on DOC/NRC Form AP– 

A and associated forms; 

(b) As required by the Additional 

Protocol, must submit location 

information described in § 75.11 of this 

chapter on DOC/NRC Form AP–1 and 

associated forms; and 

(c) Must permit verification thereof by 

the IAEA and take other action as 

necessary to implement the US/IAEA 

Safeguards Agreement, as described in 

part 75 of this chapter. 

§ 53.1660 Financial requirements. 

Sections 53.1670 through 53.1700 set 

out the requirements and procedures 

related to financial qualifications and 

related reporting requirements. 

§ 53.1670 Financial qualifications. 

Except for an electric utility applicant 

for a license to operate a commercial 

nuclear plant, an applicant for a CP, OL, 

or COL under this part must possess or 

have reasonable assurance of obtaining 

the funds necessary for the activities for 

which the permit or license is sought. 

§ 53.1680 Annual financial reports. 

With respect to any commercial 

nuclear plant of a type described in 

§ 53.020, each licensee and each holder 

of a CP must submit its annual financial 

report, including the certified financial 

statements, to the Commission, as 

specified in § 53.040, upon issuance of 

the report. However, licensees and 

holders of a CP who submit a Form 10– 

Q with the Securities and Exchange 

Commission or a Form 1 with the 

Federal Energy Regulatory Commission 

need not submit the annual financial 

report or the certified financial 

statement under this section. 

§ 53.1690 Licensee’s change of status; 

financial qualifications. 

(a) An electric utility licensee holding 

an OL or COL (including a renewed 

license) for a commercial nuclear plant, 

no later than seventy-five (75) days prior 

to ceasing to be an electric utility in any 

manner not involving a license transfer 

under § 53.1399 or § 53.1456 must 

provide the NRC with the financial 

qualifications information that would be 

required for obtaining an initial OL or 

COL under this part. The financial 

qualifications information must address 

the first full 5 years of operation after 

the date the licensee ceases to be an 

electric utility. 

(b)(1) Any holder of a license issued 

under this part must notify the 

appropriate NRC Regional 

Administrator, in writing, immediately 

following the filing of a voluntary or 

involuntary petition for bankruptcy 

under any chapter of title 11 

(Bankruptcy) of the United States Code 

by or against— 

(i) The licensee; 

(ii) An entity (as 11 U.S.C. 101(14) 

defines that term) controlling the 

licensee or listing the license or licensee 

as property of the estate; or 

(iii) An affiliate (as 11 U.S.C. 101(2) 

defines that term) of the licensee. 

(2) This notification must indicate— 

(i) The bankruptcy court in which the 

petition for bankruptcy was filed; and 

(ii) The date of the filing of the 

petition. 

§ 53.1700 Creditor regulations. 

(a) Pursuant to section 184 of the Act, 

the Commission consents, without 

individual application, to the creation of 

any mortgage, pledge, or other lien upon 

any facility not owned by the United 

States which is the subject of a license 

or upon any leasehold or other interest 

in such facility; provided— 

(1) That the rights of any creditor so 

secured may be exercised only in 

compliance with and subject to the 

same requirements and restrictions as 

would apply to the licensee pursuant to 

the provisions of the license, the Act, 

and regulations issued by the 

Commission under the Act; and 

(2) That no creditor so secured may 

take possession of the facility pursuant 

to the provisions of this section prior to 

either the issuance of a license from the 

Commission authorizing such 

possession or the transfer of the license. 


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(b) Any creditor so secured may apply 

for transfer of the license covering such 

facility by filing an application for 

transfer of the license under § 53.1570. 

The Commission will act upon such 

application under subpart I of this part. 

(c) Nothing contained in this 

regulation shall be deemed to affect the 

means of acquiring, or the priority of, 

any tax lien or other lien provided by 

law. 

(d) As used in this section— 

License includes any license under 

this part, which may be issued by the 

Commission with regard to a facility. 

Creditor includes, without implied 

limitation, the trustee under any 

mortgage, pledge or lien on a facility 

made to secure any creditor, any trustee 

or receiver of the facility appointed by 

a court of competent jurisdiction in any 

action brought for the benefit of any 

creditor secured by such mortgage, 

pledge or lien, any purchaser of such 

facility at the sale thereof upon 

foreclosure of such mortgage, pledge, or 

lien or upon exercise of any power of 

sale contained therein, or any assignee 

of any such purchaser. 

Facility includes, but is not limited to, 

a site which is the subject of an early 

site permit under this part, and a reactor 

manufactured under an ML under this 

part. 

§ 53.1710 Financial protection. 

Sections 53.1720 and 53.1730 set out 

the requirements and procedures related 

to licensees obtaining and maintaining 

insurance to cover stabilization and 

decontamination activities in the event 

of an accident and financial protection 

in accordance with part 140, ‘‘Financial 

Protection Requirements and Indemnity 

Agreements,’’ of this chapter. 

§ 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; 


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(iii) Decontamination or removal and 

disposal of internal parts, damaged fuel 

from the reactor coolant or fuel systems, 

or related process or waste systems; and 

(iv) Cleanup of the reactor coolant or 

fuel systems or related process or waste 

systems. 

(3) Following review of the licensee’s 

cleanup plan, the Director will order the 

licensee to complete all operations that 

the Director finds are necessary to 

decontaminate the reactor sufficiently to 

permit the licensee either to resume 

operation of the reactor or to apply to 

the Commission under subpart G of this 

part for authority to decommission the 

reactor and to surrender the license 

voluntarily. The Director must approve 

or disapprove, in whole or in part for 

stated reasons, the licensee’s estimate of 

cleanup costs for such operations. Such 

order may not be effective for more than 

one year, at which time it may be 

renewed. Each subsequent renewal 

order, if imposed, may be effective for 

not more than 6 months. 

(4) Of the balance of the proceeds of 

the required insurance not already 

expended to place the plant in a safe 

and stable condition under paragraph 

(b)(1) of this section, an amount 

sufficient to cover the expenses of 

completion of those decontamination 

operations that are the subject of the 

Director’s order must be dedicated to 

such use, provided that, upon 

certification to the Director of the 

amounts expended previously and from 

time to time for stabilization and 

decontamination and upon further 

certification to the Director as to the 

sufficiency of the dedicated amount 

remaining, policies of insurance may 

provide for payment to the licensee or 

other loss payees of amounts not so 

dedicated, and the licensee may proceed 

to use in parallel (and not in preference 

thereto) any insurance proceeds not so 

dedicated for other purposes. 

§ 53.1730 Financial protection 

requirements. 

Commercial nuclear plant licensees 

must satisfy the applicable provisions of 

part 140, ‘‘Financial Protection 

Requirements and Indemnity 

Agreements,’’ of this chapter. 

Subparts K and L [Reserved] 

Subpart M—Enforcement 

§ 53.9000 Violations. 

(a) The Commission may obtain an 

injunction or other court order to 

prevent a violation of the provisions 

of— 

(1) The Atomic Energy Act of 1954, as 

amended (the Act); 

(2) Title II of the Energy 

Reorganization Act of 1974, as 

amended; or 

(3) A regulation or order issued under 

those Acts. 

(b) The Commission may obtain a 

court order for the payment of a civil 

penalty imposed under Section 234 of 

the Act: 

(1) For violations of— 

(i) Sections 53, 57, 62, 63, 81, 82, 101, 

103, 104, 107, or 109 of the Act; 

(ii) Section 206 of the Energy 

Reorganization Act of 1974, as 

amended; 

(iii) Any rule, regulation, or order 

issued under the sections specified in 

paragraph (b)(1)(i) of this section; 

(iv) Any term, condition, or limitation 

of any license issued under the sections 

specified in paragraph (b)(1)(i) of this 

section. 

(2) For any violation for which a 

license may be revoked under section 

186 of the Act. 

§ 53.9010 Criminal penalties. 

(a) Section 223 of the Act provides for 

criminal sanctions for willful violation 

of, attempted violation of, or conspiracy 

to violate, any regulation issued under 

sections 161b, 161i, or 161o of the Act. 

For purposes of section 223, all the 

regulations in part 53 are issued under 

one or more of sections 161b, 161i, or 

161o, except for the sections listed in 

paragraph (b) of this section. 

(b) The regulations in 10 CFR part 53 

that are not issued under sections 161b, 

161i, or 161o for the purposes of section 

223 are as follows: §§ 53.000, 53.015, 

53.020, 53.040, 53.080, 53.090, 53.100, 

53.110, 53.120, 53.600, 53.725, 53.726, 

53.735, 53.760, 53.775, 53.790, 53.795, 

53.820, 53.910, 53.1000, 53.1050, 

53.1100, 53.1103, 53.1106, 53.1109, 

53.1112, 53.1115, 53.1118, 53.1120, 

53.1121, 53.1124, 53.1140, 53.1143, 

53.1144, 53.1146, 53.1149, 53.1155, 

53.1158, 53.1164, 53.1170, 53.1173, 

53.1176, 53.1179, 53.1188, 53.1200, 

53.1203, 53.1206, 53.1209, 53.1210, 

53.1212, 53.1215, 53.1218, 53.1221, 

53.1230, 53.1236, 53.1239, 53.1241, 

53.1242, 53.1245, 53.1248, 53.1251, 

53.1254, 52.1257, 52.1260, 53.1263, 

53.1270, 53.1273, 53.1276, 53.1279, 

53.1282, 53.1285, 53.1286, 53.1287, 

53.1288, 53.1291, 53.1293, 53.125, 

53.1300, 53.1306, 53.1309, 53.1312, 

53.1315, 53.1318, 53.1324, 53.1330, 

53.1333, 53.1336, 53.1348, 53.1360, 

53.1366, 53.1369, 53.1372, 53.1375, 

53.1381, 53.1384, 53.1387, 53.1390, 

53.1396, 53.1401, 53.1405, 53.1410, 

53.1416, 53.1419, 53.1422, 53.1425, 

53.1431, 53.1437, 53.1440, 53.1443, 

53.1452, 53.1455, 53.1456, 53.1458, 

53.1461, 53.1470, 53.1500, 53.1510, 

53.1515, 53.1520, 53.1525, 53.1530, 

53.1535, 53.1540, 53.1560, 53.1585, 

53.1590, 53.1595, 53.1600, 53.1660, 

53.1670, 53.1700, 53.1710, 53.1730, 

53.9000, 53.9010. 

PART 70—DOMESTIC LICENSING OF 

SPECIAL NUCLEAR MATERIAL 

■ 129. The authority citation for 10 CFR 

part 70 continues to read as follows: 

Authority: Atomic Energy Act of 1954, 

secs. 51, 53, 57(d), 108, 122, 161, 182, 183, 

184, 186, 187, 193, 223, 234, 274, 1701 (42 

U.S.C. 2071, 2073, 2077(d), 2138, 2152, 2201, 

2232, 2233, 2234, 2236, 2237, 2243, 2273, 

2282, 2021, 2297f); Energy Reorganization 

Act of 1974, secs. 201, 202, 206, 211 (42 

U.S.C. 5841, 5842, 5846, 5851); Nuclear 

Waste Policy Act of 1982, secs. 135, 141 (42 

U.S.C. 10155, 10161); 44 U.S.C. 3504 note. 

§ 70.20a [Amended] 

■ 130. In § 70.20a, in paragraph (b) 

remove the phrase ‘‘parts 30 through 36, 

39, 40, 50, 72, 110,’’ and add in its place 

the phrase ‘‘parts 30 through 36, 39, 40, 

50, 53, 72, 110’’. 

§ 70.22 [Amended] 

■ 131. In § 70.22, wherever it appears, 

remove the phrase ‘‘part 50’’ and add in 

its place the phrase ‘‘parst 50 or 53’’. 

■ 132. In § 70.24, revise paragraph (d) to 

read as follows: 

§ 70.24 Criticality accident requirements. 

* * * * * 

(d)(1) The requirements in paragraphs 

(a) through (c) of this section do not 

apply to a holder of a construction 

permit or operating license for a nuclear 

power reactor issued under part 50 or 

part 53 of this chapter or a combined 

license issued under part 52 or part 53 

of this chapter, if the holder complies 

with the requirements of paragraph (b) 

of 10 CFR 50.68 or paragraph (m)(2) of 

10 CFR 53.440,as applicable. 

(2) An exemption from § 70.24 held 

by a licensee who thereafter elects to 

comply with requirements of paragraph 

(b) of 10 CFR 50.68 or paragraph (m)(2) 

of 10 CFR 53.440 does not exempt that 

licensee from complying with any of the 

requirements in § 50.68 or § 53.440(m) 

of this chapter but shall be ineffective so 

long as the licensee elects to comply 

with § 50.68(b) or § 53.440(m)(2) of this 

chapter, as applicable. 

§ 70.32 [Amended] 

■ 133. In § 70.32, in paragraph (c)(1) 

introductory text, remove the phrase 

‘‘part 50 of this chapter’’ and add in its 

place the phrase ‘‘parts 50 or 53 of this 

chapter’’; and in paragraph (d) remove 

the phrase ‘‘or § 70.34 of this chapter, as 

appropriate.’’ and add in its place the 

phrase ‘‘, §§ 74.34 or 53.1510 of this 

chapter, as appropriate.’’. 


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■ 134. In § 70.50, revise paragraph (d) to 

read as follows: 

§ 70.50 Reporting requirements. 

* * * * * 

(d) The provisions of § 70.50 do not 

apply to licensees subject to §§ 50.72 or 

53.1630 of this chapter. They do apply 

to those 10 CFR parts 50 or 53 licensees 

possessing material licensed under 10 

CFR part 70 that are not subject to the 

notification requirements in §§ 50.72 or 

53.1630 of this chapter. 

PART 72—LICENSING 

REQUIREMENTS FOR THE 

INDEPENDENT STORAGE OF SPENT 

NUCLEAR FUEL AND HIGH-LEVEL 

RADIOACTIVE WASTE, AND 

REACTOR-RELATED GREATER THAN 

CLASS C WASTE 

■ 135. The authority citation for 10 CFR 

part 72 continues to read as follows: 

Authority: Atomic Energy Act of 1954, 

secs. 51, 53, 57, 62, 63, 65, 69, 81, 161, 182, 

183, 184, 186, 187, 189, 223, 234, 274 (42 

U.S.C. 2071, 2073, 2077, 2092, 2093, 2095, 

2099, 2111, 2201, 2210e, 2232, 2233, 2234, 

2236, 2237, 2238, 2273, 2282, 2021); Energy 

Reorganization Act of 1974, secs. 201, 202, 

206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); 

National Environmental Policy Act of 1969 

(42 U.S.C. 4332); Nuclear Waste Policy Act 

of 1982, secs. 117(a), 132, 133, 134, 135, 137, 

141, 145(g), 148, 218(a) (42 U.S.C. 10137(a), 

10152, 10153, 10154, 10155, 10157, 10161, 

10165(g), 10168, 10198(a)); 44 U.S.C. 3504 

note. 

■ 136. In § 72.3, revise the definition for 

‘‘Independent spent fuel storage 

installation or ISFSI’’ to read as follows: 

§ 72.3 Definitions. 

* * * * * 

Independent spent fuel storage 

installation or ISFSI means a complex 

designed and constructed for the 

interim storage of spent nuclear fuel, 

solid reactor-related GTCC waste, and 

other radioactive materials associated 

with spent fuel and reactor-related 

GTCC waste storage. An ISFSI which is 

located on the site of another facility 

licensed under this part or a facility 

licensed under part 50 or part 53 of this 

chapter and which shares common 

utilities and services with that facility or 

is physically connected with that other 

facility may still be considered 

independent. 

* * * * * 

■ 137. In § 72.30, revise paragraph (e)(5) 

to read as follows: 

§ 72.30 Financial assurance and 

recordkeeping for decommissioning. 

* * * * * 

(e) * * * 

(5) In the case of licensees who are 

issued a power reactor license under 

parts 50 or 53 of this chapter or ISFSI 

licensees who are an electric utility, as 

defined in parts 50 or 53 of this chapter, 

with a specific license issued under this 

part, the methods of §§ 50.75(b), (e), and 

(h) or 53.1010, 53.1040, 53.1045(b), and 

53.1060 of this chapter, as applicable. In 

the event that funds remaining to be 

placed into the licensee’s ISFSI 

decommissioning external sinking fund 

are no longer approved for recovery in 

rates by a competent rate making 

authority, the licensee must make 

changes to provide financial assurance 

using one or more of the methods stated 

in paragraphs (a)(1) through (4) of this 

section. 

* * * * * 

■ 138. In § 72.32, revise paragraph (c)(2) 

to read as follows: 

§ 72.32 Emergency plan. 

* * * * * 

(c) * * * 

(2)(i) Located within the exclusion 

area as defined in 10 CFR part 100, of 

a nuclear power reactor licensed for 

operation by the Commission, the 

emergency plan that meets either the 

requirements in § 50.160 of this chapter 

or the requirements in appendix E to 

part 50 of this chapter and § 50.47(b) of 

this chapter shall be deemed to satisfy 

the requirements of this section. 

(ii) Located within the exclusion area, 

as defined in 10 CFR part 53, of a 

commercial nuclear plant licensed for 

operation by the Commission, the 

emergency plan that meets either the 

requirements in § 50.160 of this chapter 

or the requirements in appendix E to 

part 50 of this chapter and § 50.47(b) of 

this chapter shall be deemed to satisfy 

the requirements of this section. 

* * * * * 

§ 72.40 [Amended] 

■ 139. In § 72.40, in paragraph (c) 

remove the phrase ‘‘under part 50 of this 

chapter,’’ and add in its place the phrase 

‘‘under parts 50 or 53 of this chapter,’’. 

■ 140. In § 72.75, revise paragraph 

(i)(1)(ii) to read as follows: 

§ 72.75 Reporting requirements for 

specific events and conditions. 

* * * * * 

(i) * * * 

(1) * * * 

(ii) Licensees issued a general license 

under § 72.210, after the licensee has 

placed spent fuel on the ISFSI storage 

pad (if the ISFSI is located inside the 

collocated protected area, for a reactor 

licensed under parts 50 or 53 of this 

chapter) or after the licensee has 

transferred spent fuel waste outside the 

reactor licensee’s protected area to the 

ISFSI storage pad (if the ISFSI is located 

outside the collocated protected area, 

for a reactor licensed under parts 50 or 

53 of this chapter). 

* * * * * 

§ 72.184 [Amended] 

■ 141. In § 72.184, in paragraph (a) 

remove the phrase ‘‘under part 50 of this 

chapter’’ and add in its place the phrase 

‘‘under parts 50 or 53 of this chapter’’. 

■ 142. Revise § 72.210 to read as 

follows: 

§ 72.210 General license issued. 

A general license is hereby issued for 

the storage of spent fuel in an 

independent spent fuel storage 

installation at power reactor sites to 

persons authorized to possess or operate 

nuclear power reactors under 10 CFR 

parts 50, 52, or 53. 

■ 143. In § 72.212, revise paragraph 

(b)(8) to read as follows: 

§ 72.212 Conditions of general license 

issued under § 72.210. 

* * * * * 

(b) * * * 

(8) Before use of the general license, 

determine whether activities related to 

storage of spent fuel under this general 

license involve a change in the facility 

Technical Specifications or require a 

license amendment for the facility 

pursuant to §§ 50.59(c) or 53.1550 of 

this chapter. Results of this 

determination must be documented in 

the evaluations made in paragraph (b)(5) 

of this section. 

* * * * * 

■ 144. In § 72.218, revise paragraphs (a) 

and (b) to read as follows: 

§ 72.218 Termination of licenses. 

(a) The notification regarding the 

program for the management of spent 

fuel at the reactor required by 

§§ 50.54(bb) or 53.1060 of this chapter 

must include a plan for removal of the 

spent fuel stored under this general 

license from the reactor site. The plan 

must show how the spent fuel will be 

managed before starting to 

decommission systems and components 

needed for moving, unloading, and 

shipping this spent fuel. 

(b) An application for termination of 

a reactor operating license issued under 

10 CFR part 50 and submitted under 

§ 50.82 of this chapter, or a combined 

license issued under 10 CFR part 52 and 

submitted under § 52.110 of this 

chapter, or a reactor operating or 

combined license under 10 CFR part 53 

and submitted under § 53.1070 of this 

chapter must contain a description of 

how the spent fuel stored under this 


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general license will be removed from 

the reactor site. 

* * * * * 

PART 73—PHYSICAL PROTECTION OF 

PLANTS AND MATERIALS 

■ 145. The authority citation for 10 CFR 

part 73 continues to read as follows: 

Authority: Atomic Energy Act of 1954, 

secs. 53, 147, 149, 161, 161A, 170D, 170E, 

170H, 170I, 223, 229, 234, 1701 (42 U.S.C. 

2073, 2167, 2169, 2201, 2201a, 2210d, 2210e, 

2210h, 2210i, 2273, 2278a, 2282, 2297f); 

Energy Reorganization Act of 1974, secs. 201, 

202 (42 U.S.C. 5841, 5842); Nuclear Waste 

Policy Act of 1982, secs. 135, 141 (42 U.S.C. 

10155, 10161); 44 U.S.C. 3504 note. 

Section 73.37(b)(2) also issued under sec. 

301, Pub. L. 96–295, 94 Stat. 789 (42 U.S.C. 

5841 note). 

■ 146. In § 73.1, revise paragraph 

(b)(1)(i) to read as follows: 

§ 73.1 [Amended] 

* * * * * 

(b) * * * 

(1) * * * 

(i) The physical protection of 

production and utilization facilities 

licensed under parts 50, 52, or 53 of this 

chapter, 

* * * * * 

■ 147. In § 73.2, revise the introductory 

text and paragraph (a) to read as follows: 

§ 73.2 Definitions. 

As used in this part: 

(a) Terms defined in parts 50, 52, 53, 

70, and 95 of this chapter have the same 

meaning when used in this part. 

* * * * * 

■ 148. In § 73.8, revise paragraph (b) to 

read as follows: 

§ 73.8 Information collection 

requirements: OMB approval. 

* * * * * 

(b) The approved information 

collection requirements contained in 

this part appear in §§ 73.5, 73.15, 73.17, 

73.20, 73.21, 73.24, 73.25, 73.26, 73.27, 

73.37, 73.40, 73.45, 73.46, 73.50, 73.54, 

73.55, 73.56, 73.57, 73.58, 73.60, 73.67, 

73.70, 73.72, 73.73, 73.74, 73.77, 73.100, 

73.110, 73.120, 73.1200, 73.1205, 

73.1210, 73.1215, and appendices B and 

C to this part. 

* * * * * 

■ 149. In § 73.50, revise the introductory 

text to read as follows: 

§ 73.50 Requirements for physical 

protection of licensed activities. 

Each licensee who is not subject to 

§ 73.51, but who possesses, uses, or 

stores formula quantities of strategic 

special nuclear material that are not 

readily separable from other radioactive 

material and which have a total external 

radiation level in excess of 1 gray (100 

rad) per hour at a distance of 1 meter 

(3.3 feet) from any accessible surfaces 

without intervening shielding other 

than at a nuclear reactor facility 

licensed under parts 50, 52, or 53 of this 

chapter, shall comply with the 

following: 

* * * * * 

■ 150. In § 73.55, revise paragraphs 

(a)(4) and (6), (i)(4)(iii), (l)(1), (l)(7)(ii), 

(p)(1)(i), (r)(2) and (r)(4)(iii) to read as 

follows: 

§ 73.55 Requirements for physical 

protection of licensed activities in nuclear 

power reactors against radiological 

sabotage. 

(a) * * * 

(4) Applicants for an operating license 

under the provisions of part 50 or part 

53 of this chapter or holders of a 

combined license under the provisions 

of part 52 or part 53 of this chapter shall 

implement the requirements of this 

section before fuel is allowed onsite 

(protected area). 

* * * * * 

(6) Applicants for an operating license 

under the provisions of part 50 or part 

53 of this chapter, or holders of a 

combined license under the provisions 

of part 52 or part 53 of this chapter that 

do not reference a standard design 

certification or reference a standard 

design certification issued after May 26, 

2009, shall meet the requirement of 

§ 73.55(i)(4)(iii). 

* * * * * 

(i) * * * 

(4) * * * 

(iii) Applicants for an operating 

license under the provisions of part 50 

of this chapter, or holders of a combined 

license under the provisions of part 52 

of this chapter, or licensees under part 

53 of this chapter that elect to 

demonstrate compliance with § 73.55, 

consistent with § 53.860(a)(2) of this 

chapter, shall construct, locate, protect, 

and equip both the central and 

secondary alarm stations to the 

standards for the central alarm station 

contained in this section. Both alarm 

stations shall be equal and redundant, 

such that all functions needed to satisfy 

the requirements of this section can be 

performed in both alarm stations. 

* * * * * 

(l) * * * 

(1) Commercial nuclear power 

reactors licensed under 10 CFR parts 50, 

52, or 53 and authorized to use special 

nuclear material in the form of MOX 

fuel assemblies containing up to 20 

weight percent PuO2 shall, in addition 

to demonstrating compliance with the 

requirements of this section, protect unirradiated MOX fuel assemblies against 

theft or diversion as described in this 

paragraph. 

* * * * * 

(7) * * * 

(ii) Additional measures for the 

physical protection of un-irradiated 

MOX fuel assemblies containing greater 

than 20 weight percent PuO2 shall be 

determined by the Commission on a 

case-by-case basis and documented 

through license amendment in 

accordance with §§ 50.90 or 53.1510 of 

this chapter. 

* * * * * 

(p) * * * 

(1) * * * 

(i) Under §§ 50.54(x) and (y) or 

53.740(h) of this chapter, the licensee 

may suspend any security measures 

under this section in an emergency 

when this action is immediately needed 

to protect the public health and safety 

and no action consistent with license 

conditions and technical specifications 

that can provide adequate or equivalent 

protection is immediately apparent. 

This suspension of security measures 

must be approved as a minimum by a 

licensed senior operator before taking 

this action. 

* * * * * 

(r) * * * 

(2) The licensee shall submit 

proposed alternative measure(s) to the 

Commission for review and approval 

under §§ 50.4 and 50.90, or §§ 53.040 

and 53.1510 of this chapter before 

implementation. 

* * * * * 

(4) * * * 

(iii) Based on comparison of the costs 

of the alternative measures to the costs 

of demonstrating compliance with the 

Commission’s requirements using the 

essential elements of §§ 50.109 or 

53.1590 of this chapter, the costs of fully 

demonstrating compliance with the 

Commission’s requirements are not 

justified by the protection that would be 

provided. 

■ 151. In § 73.56, revise paragraph (a)(3) 

to read as follows: 

§ 73.56 Personnel access authorization 

requirements for nuclear power plants. 

(a) * * * 

(3) Each applicant for an operating 

license under the provisions of part 50 

of this chapter, each holder of a 

combined license under the provisions 

of part 52 of this chapter, and applicants 

for an operating license or holders of a 

combined license under part 53 of this 

chapter that do not meet the 

requirements of § 53.860(a)(2) of this 

chapter, shall implement the 


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requirements of this section before fuel 

is allowed on site (protected area). 

* * * * * 

■ 152. In § 73.57, revise paragraph (a)(3) 

to read as follows: 

§ 73.57 Requirements for criminal history 

records checks of individuals granted 

unescorted access to a nuclear power 

facility, a non-power reactor, or access to 

Safeguards Information. 

(a) * * * 

(3) Before receiving its operating 

license under 10 CFR parts 50 or 53 or 

before the Commission makes its 

finding under §§ 52.103(g) or 53.1452(g) 

of this chapter, each applicant for a 

license to operate a nuclear power 

reactor (including an applicant for a 

combined license) or a non-power 

reactor may submit fingerprints for 

those individuals who will require 

unescorted access to the nuclear power 

facility or non-power reactor facility. 

* * * * * 

■ 153. In § 73.58, revise paragraph (a) to 

read as follows: 

§ 73.58 Safety/security interface 

requirements for nuclear power reactors. 

(a) Each operating nuclear power 

reactor licensee with a license issued 

under parts 50, 52, or 53 of this chapter 

shall comply with the requirements of 

this section. 

* * * * * 

■ 154. In § 73.67, revise paragraphs (d) 

introductory text and (f) introductory 

text to read as follows: 

§ 73.67 Licensee fixed site and in-transit 

requirements for the physical protection of 

special nuclear material of moderate and 

low strategic significance. 

* * * * * 

(d) Fixed site requirements for special 

nuclear material of moderate strategic 

significance. Each licensee who 

possesses, stores, or uses quantities and 

types of special nuclear material of 

moderate strategic significance at a fixed 

site or contiguous sites, except as 

allowed by paragraph (b)(2) of this 

section and except those who are 

licensed to operate a nuclear power 

reactor pursuant to part 50 or part 53, 

provided that the special nuclear 

material is located within a protected 

area and protected under § 73.55 or 

§ 73.100, shall: 

* * * * * 

(f) Fixed site requirements for special 

nuclear material of low strategic 

significance. Each licensee who 

possesses, stores, or uses special nuclear 

material of low strategic significance at 

a fixed site or contiguous sites, except 

those who are licensed to operate a 

nuclear power reactor pursuant to part 

50 or part 53, provided that the special 

nuclear material is located within a 

protected area and protected under 

§ 73.55 or § 73.100, shall: 

* * * * * 

■ 155. In § 73.77, revise paragraphs (a), 

(b)(1), (c)(6) and (7) to read as follows: 

§ 73.77 Cybersecurity event notifications. 

(a) Each licensee subject to the 

provisions of §§ 73.54 or 73.110 shall 

notify the NRC Headquarters Operations 

Center via the Emergency Notification 

System (ENS), under paragraph (c) of 

this section: 

(1) Within one hour after discovery of 

a cyberattack that adversely impacted: 

(i) Safety-related or important-tosafety functions, security functions, or 

emergency preparedness functions 

(including offsite communications); or 

that compromised support systems and 

equipment resulting in adverse impacts 

to safety, security, or emergency 

preparedness functions within the scope 

of § 73.54; or, 

(ii) Functions performed by digital 

assets that would prevent a postulated 

fission product release resulting in 

offsite doses exceeding the values in 

§ 53.210 of this chapter, or functions 

performed by digital assets used by the 

licensee for implementing the physical 

security requirements in § 53.860(a) of 

this chapter. 

(2) Within 4 hours: 

(i) After discovery of a cyberattack 

that could have caused an adverse 

impact to: 

(A) Safety-related or important-tosafety functions, security functions, or 

emergency preparedness functions 

(including offsite communications); or 

that could have compromised support 

systems and equipment, which if 

compromised, could have adversely 

impacted safety, security, or emergency 

preparedness functions within the scope 

of § 73.54; or, 

(B) Functions performed by digital 

assets that would prevent a postulated 

fission product release resulting in 

offsite doses exceeding the values in 

§ 53.210 of this chapter, or functions 

performed by digital assets used by the 

licensee for implementing the physical 

security requirements in § 53.860(a) of 

this chapter. 

(ii) After discovery of a suspected or 

actual cyberattack initiated by personnel 

with physical or electronic access to 

digital computer and communication 

systems and networks within the scope 

of §§ 73.54 or 73.110. 

(iii) After notification of a local, State, 

or other Federal agency (e.g., law 

enforcement, Federal Bureau of 

Investigation (FBI), etc.) of an event 

related to the licensee’s implementation 

of their cybersecurity program for digital 

computer and communication systems 

and networks within the scope of 

§§ 73.54 or 73.110 that does not 

otherwise require a notification under 

paragraph (a) of this section. 

(3) Within 8 hours after receipt or 

collection of information regarding 

observed behavior, activities, or 

statements that may indicate 

intelligence gathering or pre-operational 

planning related to a cyberattack against 

digital computer and communication 

systems and networks within the scope 

of §§ 73.54 or 73.110. 

(b) Twenty-four hour recordable 

events. (1) The licensee shall use the site 

corrective action program to record 

vulnerabilities, weaknesses, failures and 

deficiencies in their § 73.54 or § 73.110 

cybersecurity program within 24 hours 

of their discovery. 

* * * * * 

(c) * * * 

(6) Declaration of emergencies. 

Notifications made to the NRC for the 

declaration of an emergency class shall 

be performed in accordance with 

§§ 50.72 or 53.1630 of this chapter, as 

applicable. 

(7) Elimination of duplication. 

Separate notifications and reports are 

not required for events that are also 

reportable under §§ 50.72 and 50.73 or 

§§ 53.1630 and 53.1640 of this chapter. 

However, these notifications should also 

indicate the applicable § 73.77 reporting 

criteria. 

* * * * * 

■ 156. Add Subpart J consisting of 

§§ 73.100 through 73.120 to read as 

follows: 

Subpart J—Security Requirements at 

Commercial Nuclear Plants 

Sec. 

73.100 Technology-inclusive requirements 

for physical protection of licensed 

activities at commercial nuclear plants 

against radiological sabotage. 

73.110 Technology-inclusive requirements 

for protection of digital computer and 

communication systems and networks. 

73.120 Access authorization program for 

commercial nuclear plants. 

Subpart J—Security Requirements at 

Commercial Nuclear Plants 

§ 73.100 Technology-inclusive 

requirements for physical protection of 

licensed activities at commercial nuclear 

plants against radiological sabotage. 

(a) Introduction. (1) Each licensee that 

is licensed to operate a commercial 

nuclear plant under 10 CFR part 53 and 

elects to implement the requirements of 

this section must do so through its 

physical security plan, training and 

qualification plan, safeguards 


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contingency plan, and cybersecurity 

plan, referred to collectively hereafter as 

‘‘security plans,’’ before initial fuel load 

into the reactor (or, for a fueled 

manufactured reactor, before initiating 

the physical removal of any one of the 

independent physical mechanisms to 

prevent criticality required under 

§ 53.620(d)(1) of this chapter). 

(2) The security plans must identify, 

describe, and account for site-specific 

conditions that affect the licensee’s 

capability to satisfy the requirements of 

this section. 

(b) General performance objective and 

requirements. (1) The licensee must 

establish, implement, and maintain a 

physical protection program and a 

security organization, which will have 

as their objective to provide reasonable 

assurance that activities involving 

special nuclear material are not inimical 

to the common defense and security and 

do not constitute an unreasonable risk 

to the public health and safety. 

(2) To satisfy the general performance 

objective of paragraph (b)(1) of this 

section, the physical protection program 

must protect against the design basis 

threat of radiological sabotage as stated 

in § 73.1. Specifically, the licensee 

must— 

(i) Ensure that the physical protection 

program capabilities to protect against 

the design basis threat of radiological 

sabotage are maintained at all times; and 

(ii) Provide defense in depth in 

achieving performance requirements 

through the integration of engineered 

systems, administrative controls, and 

management measures. 

(3) The physical protection program 

must be designed and implemented to 

achieve and maintain the reliability and 

availability of structures, systems, and 

components (SSCs) required for 

demonstrating compliance with the 

following performance requirements at 

all times: 

(i) Intrusion detection. The licensee 

must be capable of detecting attempted 

and actual unauthorized access to 

interior and exterior areas containing 

SSCs needed to implement safety and 

security functions. 

(ii) Intrusion assessment. The licensee 

must be capable of timely assessment 

for determining the cause of a detected 

intrusion. 

(iii) Security communication. The 

licensee must be capable of continuous 

security communications. 

Communication systems must account 

for design basis threats that can 

interrupt or interfere with continuity or 

integrity of communications. 

(iv) Security response. The physical 

protection program must be designed to 

provide timely security response to 

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 securityrelated systems, and components from 

explosive effects. 

(vi) Access control portals. The 

licensee must be capable of detecting 

and denying unauthorized access to 

persons and pass-through of contraband 

materials (e.g., weapons, incendiaries, 

explosives) to protected areas. 

(4) The licensee must meet the 

requirements related to target sets in 

§ 73.55(f). 

(5) The licensee must identify and 

analyze site-specific conditions, 

including target sets, that may affect the 

physical protection program needed to 

implement the requirements of this 

section. The licensee must account for 

these conditions in demonstrating 

compliance with the requirements of 

this section. 

(6) The licensee must establish, 

implement, and maintain a performance 

evaluation program to assess the 

effectiveness of the licensee’s 

implementation of the physical 

protection program to protect against 

the design basis threat of radiological 

sabotage. 

(7) The licensee must establish, 

implement, and maintain an access 

authorization program under § 73.56 

and must describe the program in the 

physical security plan. 

(8) The licensee must establish, 

implement, and maintain a 

cybersecurity program under §§ 73.54 or 

73.110 and must describe the program 

in the cybersecurity plan. 

(9) The licensee must establish, 

implement, and maintain an insider 

mitigation program and must describe 

the program in the physical security 

plan. 

(i) The insider mitigation program 

must monitor the initial and continuing 

trustworthiness and reliability of 

individuals granted or retaining 

unescorted access or unescorted access 

authorization to a protected or vital 

area, and implement defense-in-depth 

methodologies to minimize the potential 

for an insider (active, passive, or both) 


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to adversely affect, either directly or 

indirectly, the licensee’s capability to 

protect against radiological sabotage. 

(ii) The insider mitigation program 

must integrate elements of— 

(A) The access authorization program 

under § 73.56; 

(B) The fitness-for-duty program 

under 10 CFR part 26; 

(C) The cybersecurity program under 

§§ 73.54 or 73.110; and 

(D) The physical protection program 

under this section. 

(10) The licensee must have the 

capability to track, trend, correct, and 

prevent recurrence of failures and 

deficiencies in the implementation of 

the requirements of this section. 

(11) Implementation of security plans 

and associated procedures must be 

coordinated with other onsite plans and 

procedures to preclude conflict during 

both normal and emergency conditions 

and ensure the adequate management of 

the safety and security interface. 

(12)(i) The licensee must ensure that 

the firearms background check 

requirements of § 73.17 of this part are 

met for all members of the security 

organization whose official duties 

require access to covered weapons or 

who inventory enhanced weapons. 

(ii) The provisions of this paragraph 

are only applicable to licensees subject 

to this section that are also subject to the 

firearms background check provisions of 

§ 73.17 of this part. 

(c) Security organization. The licensee 

must establish and maintain a security 

organization that is staffed, trained, 

qualified, and equipped to implement 

the physical protection program under 

the requirements of this section. 

(1) The licensee must establish a 

management system for maintaining and 

implementing security policies and 

procedures to implement the 

requirements of this section and the 

security plans. 

(2) Implementing procedures must 

document the conduct of security 

operations, security design and 

configuration controls, maintenance, 

training and qualification, and 

contingency responses. 

(3) The licensee must— 

(i) Establish a process for the approval 

of designs, policies, processes, and 

procedures and changes by the 

individual with overall responsibility 

for the physical protection program; and 

(ii) Ensure that revisions and changes 

to the physical protection program and 

implementing policies, processes, and 

procedures satisfy the requirements of 

this section. 

(4) The licensee must retain, in 

accordance with § 73.70, all analyses, 

assessments, calculations, and 

descriptions of the technical basis for 

demonstrating compliance with the 

performance requirements of 

§ 73.100(b). The licensee must protect 

these records in accordance with the 

requirements for protecting safeguards 

information in §§ 73.21 and 73.22. 

(5) The licensee may not permit any 

individual to implement any part of the 

physical protection program unless the 

individual has been trained, equipped, 

and qualified to perform their assigned 

duties and responsibilities in 

accordance with the training and 

qualification plan. 

(d) Search requirements. The licensee 

must establish and implement searches 

of individuals, vehicles, and materials 

to detect and prevent the introduction 

into the protected area of firearms, 

explosives, incendiary devices, or other 

items and material which could be used 

to commit radiological sabotage. 

(e) Training and qualification 

program. The licensee must establish 

and maintain a training and 

qualification program that ensures 

personnel who are responsible for the 

physical protection of the facility 

against radiological sabotage are able to 

effectively perform their assigned 

security-related job duties for 

implementing the requirements of this 

section and must describe the program 

in the training and qualification plan. 

(f) Security reviews. The licensee must 

establish and implement security 

reviews to assess the effectiveness of the 

implementation of the physical 

protection program. Security reviews 

must be performed by individuals 

independent of those personnel 

responsible for program management 

and any individual who has direct 

responsibility for implementing the 

onsite physical protection program. 

(1) The licensee must review each 

element of the physical protection 

program at a frequency commensurate 

with the importance or significance to 

safety of plant operations to ensure 

timely identification and documentation 

of vulnerabilities, improvements, and 

corrective actions. The objective of these 

reviews must be maintaining effective 

implementation of the engineered and 

administrative controls required to 

achieve the physical protection program 

functions and the management system 

required to implement programs and 

requirements in this section. 

(2) The licensee must establish and 

perform self-assessments to ensure the 

effective implementation of the physical 

protection program functions of 

detection, assessment, communication, 

delay, and interdiction and 

neutralization to protect against the 

design basis threat of radiological 

sabotage. The licensee must perform 

design verification and assessments of 

the capabilities of active and passive 

engineering systems relied on to protect 

against the design basis threat. 

(3) Reviews of the security program 

must include, but are not limited to, an 

audit of the effectiveness of the physical 

protection program, security plans, 

implementing procedures, cybersecurity 

programs, safety/security interface 

activities, the testing, maintenance, and 

calibration program, and response 

commitments by local, State, and 

Federal law enforcement authorities. 

(4) The results and recommendations 

of the onsite physical protection 

program reviews, management’s 

findings regarding program 

effectiveness, and any actions taken as 

a result of recommendations from prior 

program reviews, must be documented 

in a report and must be maintained in 

an auditable form and available for 

inspection. 

(g) Performance evaluation. Licensee 

performance evaluations must include 

methods appropriate and necessary to 

assess, test, and challenge the 

integration of the physical protection 

program’s functions to protect against 

the design basis threat, including 

measures to protect against cyberattack 

and engineered systems designed to 

protect against the design basis threat 

standalone ground vehicle bomb attack. 

(1) The licensee must establish the 

frequencies for performance evaluations 

commensurate with the security 

significance of the physical protection 

program. 

(2) The licensee must document 

processes and procedures for 

implementing the performance 

evaluations. The licensee must maintain 

records, including results, findings, and 

corrective actions identified during the 

performance evaluations. 

(h) Maintenance, testing, and 

calibration and corrective actions. (1) 

The licensee must ensure that security 

SSCs, including supporting systems, are 

inspected, tested, and calibrated for 

operability and performance at intervals 

necessary and sufficient to meet the 

requirements of this section. 

(2) The licensee must implement 

corrective actions to ensure resolution 

of identified vulnerabilities and 

deficiencies to meet the requirements of 

this section. 

(3) The licensee must establish and 

implement timely compensatory 

measures for degraded or inoperable 

security SSCs to meet the requirements 

of this section. Compensatory measures 

must provide a level of protection that 

is equivalent to the protection that was 

provided prior to the degradation or 


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inoperability of the security structures, 

systems, or components. 

(4) The licensee must document 

processes and procedures and maintain 

records for implementing the corrective 

actions, compensatory measures, and 

maintenance, inspection, testing, and 

calibration of security SSCs. 

(i) Suspension of security measures. 

(1) The licensee may suspend 

implementation of affected 

requirements of this section in 

accordance with § 53.740(h) of this 

chapter under the following conditions: 

(i) In an emergency, when action is 

immediately needed to protect the 

public health and safety; and 

(ii) During severe weather, when the 

suspension of affected security 

measures is immediately needed to 

protect the personal health and safety of 

personnel. 

(2) Suspended security measures must 

be reinstated as soon as conditions 

permit. 

(3) The suspension of security 

measures must be reported and 

documented in accordance with the 

provisions of §§ 73.1200 and 73.1205. 

(j) Records. (1) The Commission may 

inspect, copy, retain, and remove all 

reports, records, and documents 

required to be kept by Commission 

regulations, orders, or license 

conditions, whether the reports, records, 

and documents are kept by the licensee 

or a contractor. 

(2) The licensee must maintain all 

records required to be kept by 

Commission regulations, orders, or 

license conditions, until the 

Commission terminates the license for 

which the records were developed and 

must maintain superseded portions of 

these records for at least 3 years after the 

record is superseded, unless otherwise 

specified by the Commission. 

(3) If a contracted security force is 

used to implement the onsite physical 

protection program, the licensee’s 

written agreement with the contractor 

must be retained by the licensee as a 

record for the duration of the contract. 

(4) Review and audit reports must be 

available for inspection, for a period of 

3 years. 

§ 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 

§ 73.1. The cybersecurity program must 

provide reasonable assurance that 

digital computer and communication 

systems and networks are adequately 

protected against cyberattacks that are 

capable of causing the following 

consequences: 

(1) Adversely impacting the functions 

performed by digital assets that would 

prevent a postulated fission product 

release resulting in offsite doses 

exceeding the values in § 53.210 of this 

chapter. 

(2) Adversely impacting the functions 

performed by digital assets used by the 

licensee for implementing the physical 

security requirements in § 53.860(a) of 

this chapter. 

(b) To protect digital computer and 

communication systems and networks 

associated with the functions described 

in paragraphs (a)(1) and (2), the licensee 

must— 

(1) Analyze the potential 

consequences resulting from 

cyberattacks on digital computer and 

communication systems and networks 

and identify those assets that must be 

protected to demonstrate compliance 

with paragraph (a) of this section; and 

(2) Implement the cybersecurity 

program in accordance with paragraph 

(d) of this section. 

(c) The licensee must comply with the 

requirements in § 73.54(a)(2) for the 

systems and networks identified in 

paragraph (b)(1) of this section in a 

manner that is commensurate with the 

potential consequences resulting from 

cyberattacks. 

(d) The cybersecurity program must 

be designed in a manner that is 

commensurate with the potential 

consequences resulting from 

cyberattacks through the following 

steps: 

(1) Implement security controls to 

protect the assets identified under 

paragraph (b)(1) of this section from 

cyberattacks, commensurate with their 

safety and security significance; 

(2) Apply and maintain defense-indepth protective strategies to ensure the 

capability to detect, delay, respond to, 

and recover from cyberattacks capable 

of causing the consequences identified 

in paragraph (a) of this section; 

(3) Mitigate the adverse effects of 

cyberattacks capable of causing the 

consequences identified in paragraph (a) 

of this section; and 

(4) Ensure that the functions of 

protected assets identified under 

paragraph (b)(1) of this section are not 

adversely impacted due to cyberattacks. 

(e) The licensee must implement the 

following requirements in a manner that 

is commensurate with the potential 

consequences resulting from 

cyberattacks: 

(1) As part of the cybersecurity 

program, the licensee must comply with 

the requirements in § 73.54(d)(1), (2), 

and (4), and must ensure that 

modifications to assets, identified under 

paragraph (b)(1) of this section are 

evaluated before implementation to 

ensure that the cybersecurity 

performance objectives identified in 

paragraph (a) of this section are 

maintained. 

(2) The licensee must establish, 

implement, and maintain a 

cybersecurity plan that implements the 

cybersecurity program requirements of 

this section. 

(i) The cybersecurity plan must 

describe how the requirements of this 

section will be implemented and must 

account for the site-specific conditions 

that affect implementation. 

(ii) The cybersecurity plan must 

include measures for incident response 

and recovery for cyberattacks. The 

cybersecurity plan must include the 

analysis identified under paragraph 

(b)(1) of this section and describe how 

the licensee will— 

(A) Apply and maintain defense-indepth protective strategies as required 

in paragraph (d)(2) of this section; 

(B) Maintain the capability for timely 

detection and response to cyberattacks; 

(C) Mitigate the consequences of 

cyberattacks; 

(D) Correct exploited vulnerabilities; 

and 

(E) Restore affected systems, 

networks, and/or equipment affected by 

cyberattacks. 

(3) The licensee must develop and 

maintain written policies and 

implementing procedures to implement 

the cybersecurity plan. Policies, 

implementing procedures, and other 

supporting technical information used 

by the licensee need not be submitted 

for Commission review and approval as 

part of the cybersecurity plan but are 

subject to inspection by NRC staff on a 

periodic basis. 

(4) The licensee must establish and 

implement cybersecurity reviews to 

assess the effectiveness of the 

implementation of the cybersecurity 

program. 

(i) The licensee must review each 

element of the cybersecurity program at 

a frequency commensurate with the 

importance or significance to safety of 

plant operations to ensure timely 

identification and documentation of 

vulnerabilities, improvements, and 

corrective actions. 

(ii) Cybersecurity reviews must be 

performed by individuals independent 


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of those personnel responsible for 

program management and any 

individual who has direct responsibility 

for implementing the cybersecurity 

program. 

(iii) The licensee must establish and 

perform self-assessments to ensure the 

effective implementation of the 

cybersecurity program. 

(iv) The results and recommendations 

of the cybersecurity program reviews, 

management’s findings regarding 

program effectiveness, and any actions 

taken as a result of recommendations 

from prior program reviews, must be 

documented in a report and must be 

maintained in an auditable form and 

available for inspection. 

(5) The licensee must retain all 

records and supporting technical 

documentation required to demonstrate 

compliance with the requirements of 

this section as a record until the 

Commission terminates the license for 

which the records were developed and 

must maintain superseded portions of 

these records for at least three (3) years 

after the record is superseded, unless 

otherwise specified by the Commission. 

§ 73.120 Access authorization program for 

commercial nuclear plants. 

(a) Introduction and scope. Each 

applicant for an operating license or a 

holder of a combined license under 10 

CFR part 53 must establish, maintain, 

and implement an access authorization 

program before initial fuel load into the 

reactor (or, for a fueled manufactured 

reactor, before initiating the physical 

removal of any one of the independent 

physical mechanisms to prevent 

criticality required under § 53.620(d)(1) 

of this chapter). The requirements in 

this section apply to licensees satisfying 

the criterion in § 53.860(a)(2)(i) of this 

chapter. 

(b) Applicability. (1) The following 

individuals must be subject to an access 

authorization program under this 

section: 

(i) Any individual to whom a licensee 

intends to grant unescorted access to a 

commercial nuclear plant protected 

area, vital area, or controlled access area 

where licensed material is used or 

stored; 

(ii) Any individual whose duties and 

responsibilities permit the individual to 

take actions by electronic means, either 

on site or remotely, that could adversely 

impact the licensee’s or applicant’s 

operational safety, security, or 

emergency preparedness; 

(iii) Any individual who has 

responsibilities for implementing a 

licensee’s or applicant’s protective 

strategy, including armed security force 

officers, alarm station operators, and 

tactical response team leaders but not 

including Federal, State, or local law 

enforcement personnel; and 

(iv) The licensee or applicant access 

authorization program reviewing official 

or contractor or vendor access 

authorization program reviewers. 

(2) The licensee or applicant may 

subject other individuals, including 

employees of a contractor or a vendor 

who are designated in access 

authorization program procedures, to an 

access authorization program that 

demonstrates compliance with the 

requirements of this section. 

(c) General performance objectives 

and requirements. Each licensee’s or 

applicant’s access authorization 

program under this section must 

demonstrate that the individuals who 

are specified in paragraph (b) of this 

section are trustworthy and reliable, 

such that they do not constitute an 

unreasonable risk to public health and 

safety or the common defense and 

security. The licensee’s access 

authorization program must maintain 

the capabilities for demonstrating 

compliance with the following 

performance requirements: 

(1) Background investigation. (i)(A) 

Licensees and applicants must ensure 

that any individual seeking initial 

unescorted access or to maintain 

unescorted access is subject to a 

background investigation. 

(B) Background investigations must 

include the program elements contained 

under § 37.25 of this chapter and must 

also include a credit history evaluation. 

(ii) Background investigations must 

include fingerprinting and an FBI 

identification and criminal history 

records check in accordance with 

§ 37.27 of this chapter. 

(iii) Licensees must have the informed 

and signed consent of the subject 

individual to initiate a background 

investigation. This consent must 

include authorization to share personal 

information with other individuals or 

organizations as necessary to complete 

the background investigation. A signed 

consent must be obtained prior to any 

reinvestigation. The subject individual 

may withdraw his or her consent at any 

time. Licensees must inform the 

individual that— 

(A) If an individual withdraws his or 

her consent, the licensee may not 

initiate any elements of the background 

investigation that were not in progress 

at the time the individual withdrew his 

or her consent; and 

(B) The withdrawal of consent for the 

background investigation is sufficient 

cause for denial or termination of 

unescorted access authorization. 

(2) Behavioral observation. Licensees, 

applicants, contractors, and vendors 

must ensure the access authorization 

program includes provisions that the 

individuals specified in paragraph (b) of 

this section are subject to behavioral 

observation. 

(i) Each person subject to behavioral 

observation must communicate to the 

licensee or applicant observed behaviors 

or activities of individuals that may 

constitute an unreasonable risk to the 

health and safety of the public and 

common defense and security. 

(ii) Behavioral observation must 

include visual observation, in person or 

remotely by video, to detect and 

promptly report to plant supervision 

any concerns arising from behavioral 

observation, including, but not limited 

to, concerns related to any questionable 

behavior patterns or activities of others. 

(3) Self-reporting of legal actions. 

Licensees or applicants must inform 

personnel who are granted and who 

maintain unescorted access of their 

responsibilities to self-report to plant 

supervision legal actions taken by a law 

enforcement authority or court of law 

against the individual that could result 

in incarceration or a court order or that 

requires a court appearance, including 

but not limited to an arrest, an 

indictment, the filing of charges, or a 

conviction, but excluding minor civil 

actions or misdemeanors such as 

parking violations or speeding tickets, 

for any individual who has applied for 

unescorted access or who maintains 

unescorted access. 

(4) Unescorted access. Licensees or 

applicants must grant unescorted access 

only after the licensee has verified an 

individual is trustworthy and reliable. A 

list of persons currently approved for 

unescorted access to a protected area, 

vital area, or controlled access area must 

be maintained at all times. Unescorted 

access determinations must be reviewed 

annually by the reviewing official. 

Licensees and applicants must complete 

an FBI criminal history record check 

update for each individual maintaining 

unescorted access, within 10 years of 

the last review. 

(5) Termination of unescorted access. 

Licensees and applicants must promptly 

terminate unescorted access when this 

access is no longer required or a 

reviewing official determines an 

individual is no longer trustworthy and 

reliable in accordance with this section. 

(6) Determination basis for access. (i) 

The licensee’s or applicant’s reviewing 

official must determine whether to 

permit, deny, unfavorably terminate, 

maintain, or administratively withdraw 

an individual’s unescorted access based 

on an evaluation of all of the 


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information collected to demonstrate 

compliance with the requirements of 

this section. 

(ii) Licensees and applicants must 

provide individuals subject to this 

section, prior to any final adverse 

determination, the right to complete, 

correct, and explain information 

obtained as a result of the licensee’s 

background investigation pursuant to 

§ 37.23(g) of this chapter. 

(iii) The licensee’s or applicant’s 

reviewing officials are the only 

individuals authorized to make 

unescorted access determination 

decisions. Each licensee or applicant 

must name one or more individuals to 

be reviewing officials pursuant to the 

requirements of § 37.23(b)(2) of this 

chapter. 

(7) Review procedures. Review 

procedures must be established in 

accordance with § 37.23(f) of this 

chapter, to include provisions for the 

notification in writing of individuals 

who are denied unescorted access or 

who are unfavorably terminated. 

(8) Protection of information. 

Licensees, applicants, contractors, or 

vendors must establish and maintain a 

system of files and procedures in 

accordance with § 37.31 of this chapter, 

to ensure personal information is not 

disclosed to unauthorized persons. 

(9) Access authorization reviews and 

corrective action. Licensees and 

applicants must develop, implement, 

and maintain procedures for conduct of 

access authorization reviews and 

corrective actions in accordance with 

§ 37.33 of this chapter to ensure the 

continuing effectiveness of the access 

authorization program and to ensure 

that the access authorization program 

and program elements are in 

compliance with the requirements of 

this section. Each licensee and applicant 

must be responsible for the continuing 

effectiveness of the access authorization 

program, including access authorization 

program elements that are provided by 

the contractors or vendors, and the 

access authorization programs of any of 

the contractors or vendors that are 

accepted by the licensee or applicant. 

(10) Records. Licensees, applicants, 

and contractors or vendors must 

document the processes and procedures 

for maintaining records used or created 

to establish an individual’s 

trustworthiness and reliability or to 

document access determinations. 

Licensees, applicants, and contractor or 

vendors must— 

(i) Retain documentation regarding 

the trustworthiness and reliability of 

individual employees for 3 years from 

the date the individual no longer 

requires unescorted access; 

(ii) Retain a copy of the current access 

authorization program procedures as a 

record for 3 years after the procedure is 

no longer needed. If any portion of the 

procedure is superseded, retain the 

superseded material for 3 years after the 

record is superseded; and 

(iii) Retain the list of persons 

approved for unescorted access for 3 

years after the list is superseded or 

replaced. Records maintained in any 

database(s) must be available for NRC 

review. 

■ 157. In § 73.1200, revise paragraphs 

(a) introductory text, (c)(1) introductory 

text, (e)(1) introductory text, (e)(3) and 

(4), (g)(1) introductory text, (o)(5)(i) and 

(o)(6)(i), (r) and (s) to read as follows: 

§ 73.1200 Notification of physical security 

events. 

(a) 15-minute notifications—facilities. 

Each licensee subject to the provisions 

of § 73.20, § 73.45, § 73.46, § 73.51, 

§ 73.55, or § 73.100 must notify the NRC 

Headquarters Operations Center, as soon 

as possible but within 15 minutes 

after— 

* * * * * 

(c) * * * 

(1) Each licensee subject to the 

provisions of §§ 73.20, 73.45, 73.46, 

73.50, 73.51, 73.55, 73.60, 73.67, or 

73.100 must notify the NRC 

Headquarters Operations Center as soon 

as possible but no later than 1 hour after 

the time of discovery of the following 

significant facility security events 

involving— 

* * * * * 

(e) * * * 

(1) Each licensee subject to the 

provisions of §§ 73.20, 73.45, 73.46, 

73.50, 73.51, 73.55, 73.60, 73.67, or 

73.100 must notify the NRC 

Headquarters Operations Center within 

4 hours after time of discovery of the 

following facility security events 

involving— 

* * * * * 

(3)(i) An event involving a law 

enforcement response to the facility that 

could reasonably be expected to result 

in public or media inquiries and that 

does not otherwise require a notification 

under paragraphs (a) through (h) of this 

section, or in other NRC regulations 

such as § 50.72(b)(2)(xi) or 

§ 53.1630(b)(2)(v) of this chapter. 

(ii) As an exemption, licensees need 

not report law enforcement responses to 

minor incidents, such as traffic 

accidents. 

(4) For licensees subject to the 

provisions of § 73.55 or § 73.100 of this 

part, an event involving the licensee’s 

suspension of security measures. 

* * * * * 

(g) * * * 

(1) Each licensee subject to the 

provisions of § 73.20, § 73.45, § 73.46, 

§ 73.50, § 73.51, § 73.55, § 73.60, § 73.67, 

or § 73.100 must notify the NRC 

Headquarters Operations Center within 

8 hours after time of discovery of the 

following facility security program 

failures involving— 

* * * * * 

(o) * * * 

(5) * * * 

(i) Licensees must establish the 

requested continuous communications 

channel once the licensee has 

completed other required notifications 

under this section, § 50.72 of this 

chapter, appendix E to part 50 of this 

chapter, § 53.1630 of this chapter, 

§ 70.50 of this chapter; or § 72.75 of this 

chapter; as appropriate. 

* * * * * 

(6) * * * 

(i) Licensees must establish the 

requested continuous communications 

channel once the licensee or the 

movement control center has completed 

other required notifications under this 

section, § 50.72 of this chapter, 

appendix E to part 50 of this chapter, 

§ 53.1630 of this chapter, § 70.50 of this 

chapter; § 72.75 of this chapter; or 

requested assistance from the LLEA, as 

appropriate. 

* * * * * 

(r) Declaration of emergencies. 

Licensees notifying the NRC of the 

declaration of an emergency class must 

do so in accordance with §§ 50.72, 

53.1630, 63.73, 70.50, and 72.75 of this 

chapter, as applicable. 

(s) Elimination of duplication. 

Licensees with notification obligations 

under paragraphs (a) through (h), (m), 

and (n) of this section and §§ 50.72, 

53.1630, 63.73, 70.50, and 72.75 of this 

chapter may notify the NRC of events in 

a single communication. This 

communication must identify each 

regulation under which the licensee is 

reporting. 

* * * * * 

■ 158. In § 73.1205, revise paragraph 

(b)(2) to read as follows: 

§ 73.1205 Written follow-up reports of 

physical security events. 

* * * * * 

(b) * * * 

(2)(i) Licensees subject to § 50.73 or 

§ 53.1640 of this chapter must prepare 

the written follow-up report on NRC 

Form 366. 

(ii) Licensees not subject to § 50.73 or 

§ 53.1640 of this chapter must prepare 

the written follow-up report in a letter 

format. 

* * * * * 


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■ 159. In § 73.1210, revise paragraphs 

(a)(1) and (b)(3)(i) to read as follows: 

§ 73.1210 Recordkeeping of physical 

security events. 

(a) * * * 

(1) Licensees with facilities or 

shipment activities subject to the 

provisions of § 73.20, § 73.25, § 73.26, 

§ 73.27, § 73.37, § 73.45, § 73.46, § 73.50, 

§ 73.51, § 73.55, § 73.60, § 73.67, or 

§ 73.100, must record the physical 

security events and conditions adverse 

to security that are specified in 

paragraphs (c) through (f) of this section. 

* * * * * 

(b) * * * 

(3)(i) Licensees must record these 

physical security events and conditions 

adverse to security in either a standalone safeguards event log or as part of 

the licensee’s corrective action program, 

as specified under the applicable quality 

assurance program provisions of parts 

50, 52, 53, 60, 63, 70, and 72 of this 

chapter, or both. 

* * * * * 

■ 160. In § 73.1215, revise paragraph 

(d)(1) introductory text to read as 

follows: 

§ 73.1215 Suspicious activity reports. 

* * * * * 

(d) * * * 

(1) For licensees subject to the 

provisions of §§ 73.20, 73.45, 73.46, 

73.50, 73.51, 73.55, 73.60, 73.67, or 

73.100, the licensees must report 

activities they assess are suspicious. 

Examples include, but are not limited 

to, the following: 

* * * * * 

■ 161. In appendix B to part 73, revise 

Definitions introductory text to read as 

follows: 

Appendix B to Part 73—General 

Criteria for Security Personnel 

* * * * * 

Definitions 

Terms defined in parts 50, 53, 70, and 73 

of this chapter have the same meaning when 

used in this appendix. 

* * * * * 

PART 74—MATERIAL CONTROL AND 

ACCOUNTING OF SPECIAL NUCLEAR 

MATERIAL 

■ 162. The authority citation for 10 CFR 

part 74 continues to read as follows: 

Authority: Atomic Energy Act of 1954, 

secs. 53, 57, 161, 182, 223, 234, 1701 (42 

U.S.C. 2073, 2077, 2201, 2232, 2273, 2282, 

2297f); Energy Reorganization Act of 1974, 

secs. 201, 202 (42 U.S.C. 5841, 5842); 44 

U.S.C. 3504 note. 

■ 163. In § 74.31, revise paragraph (a) 

introductory text to read as follows: 

§ 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: 

* * * * * 

■ 164. In § 74.41, revise paragraph (a) 

introductory text to read as follows: 

§ 74.41 Nuclear material control and 

accounting for special nuclear material of 

moderate strategic significance. 

(a) General performance objectives. 

Each licensee who is authorized to 

possess special nuclear material (SNM) 

of moderate strategic significance or 

SNM in a quantity exceeding one 

effective kilogram of strategic special 

nuclear material in irradiated fuel 

reprocessing operations other than as 

sealed sources and to use this material 

at any site other than a nuclear reactor 

licensed pursuant to parts 50 or 53 of 

this chapter; or as reactor irradiated 

fuels involved in research, 

development, and evaluation programs 

in facilities other than irradiated fuel 

reprocessing plants; or an operation 

involved with waste disposal, shall 

establish, implement, and maintain a 

Commission-approved material control 

and accounting (MC&A) system that will 

achieve the following performance 

objectives: 

* * * * * 

■ 165. In § 74.51, revise paragraph (a) 

introductory text to read as follows: 

§ 74.51 Nuclear material control and 

accounting for strategic special nuclear 

material. 

(a) General performance objectives. 

Each licensee who is authorized to 

possess five or more formula kilograms 

of strategic special nuclear material 

(SSNM) and to use such material at any 

site, other than a nuclear reactor 

licensed pursuant to parts 50 or 53 of 

this chapter, an irradiated fuel 

reprocessing plant, an operation 

involved with waste disposal, or an 

independent spent fuel storage facility 

licensed pursuant to part 72 of this 

chapter shall establish, implement, and 

maintain a Commission-approved 

material control and accounting (MC&A) 

system that will achieve the following 

objectives: 

* * * * * 

PART 75—SAFEGUARDS ON 

NUCLEAR MATERIAL— 

IMPLEMENTATION OF SAFEGUARDS 

AGREEMENTS BETWEEN THE UNITED 

STATES AND THE INTERNATIONAL 

ATOMIC ENERGY AGENCY 

■ 166. The authority citation for 10 CFR 

part 75 continues to read as follows: 

Authority: Atomic Energy Act of 1954, 

secs. 53, 63, 103, 104, 122, 161, 223, 234, 

1701 (42 U.S.C. 2073, 2093, 2133, 2134, 2152, 

2201, 2273, 2282, 2297f); Energy 

Reorganization Act of 1974, sec. 201 (42 

U.S.C. 5841); Nuclear Waste Policy Act of 

1982, secs. 135, 141 (42 U.S.C. 10155, 10161); 

44 U.S.C. 3504 note. 

■ 167. In § 75.4, revise the introductory 

text and paragraph (6) of the definition 

for ‘‘Facility’’ to read as follows: 

§ 75.4 Definitions. 

* * * * * 

Unless otherwise defined in this 

section, the terms defined in §§ 40.4, 

50.2, 53.020, and 70.4 of this chapter 

have the same meaning when used in 

this part. 

* * * * * 

Facility means: 

* * * * * 

(6) Any plant or location where the 

possession of more than 1 effective 

kilogram of nuclear material is licensed 

pursuant to 10 CFR part 40, 50, 53, 60, 

61, 63, 70, 72, 76, or 150 of this chapter 

or an Agreement State license. 

* * * * * 

PART 95—FACILITY SECURITY 

CLEARANCE AND SAFEGUARDING 

OF NATIONAL SECURITY 

INFORMATION AND RESTRICTED 

DATA 

■ 168. The authority citation for 10 CFR 

part 95 continues to read as follows: 

Authority: Atomic Energy Act of 1954, 

secs. 145, 161, 223, 234 (42 U.S.C. 2165, 

2201, 2273, 2282); Energy Reorganization Act 

of 1974, sec. 201 (42 U.S.C. 5841); 44 U.S.C. 

3504 note; E.O. 10865, as amended, 25 FR 

1583, 3 CFR, 1959–1963 Comp., p. 398; E.O. 

12829, 58 FR 3479, 3 CFR, 1993 Comp., p. 

570; E.O. 12968, 60 FR 40245, 3 CFR, 1995 

Comp., p. 391; E.O. 13526, 75 FR 707, 3 CFR, 

2009 Comp., p. 298. 

■ 169. In § 95.5, revise the definition for 

‘‘License’’ to read as follows: 

§ 95.5 Definitions. 

* * * * * 

License means a license issued under 

10 CFR part 50, 52, 53, 54, 60, 63, 70, 

or 72. 

* * * * * 


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§ 95.39 [Amended] 

■ 170. In § 95.39(a), remove ‘‘part 52’’ 

and add in its place ‘‘parts 52 or 53.’’ 

PART 140—FINANCIAL PROTECTION 

REQUIREMENTS AND INDEMNITY 

AGREEMENTS 

■ 171. The authority citation for 10 CFR 

part 140 continues to read as follows: 

Authority: Atomic Energy Act of 1954, 

secs. 161, 170, 223, 234 (42 U.S.C. 2201, 

2210, 2273, 2282); Energy Reorganization Act 

of 1974, secs. 201, 202 (42 U.S.C. 5841, 

5842); 44 U.S.C. 3504 note. 

■ 172. In § 140.2, revise paragraphs 

(a)(1) and (2) to read as follows: 

§ 140.2 Scope. 

(a) * * * 

(1) To each person who is an 

applicant for or holder of a license 

issued under 10 CFR part 50, 52, 53, or 

54 to operate a nuclear reactor, and 

(2) With respect to an extraordinary 

nuclear occurrence, to each person who 

is an applicant for or holder of a license 

to operate a production facility or a 

utilization facility (including an 

operating license issued under part 50 

or part 53 of this chapter and a 

combined license under part 52 or part 

53 of this chapter), and to other persons 

indemnified with respect to the 

involved facilities. 

* * * * * 

■ 173. Revise § 140.10 to read as 

follows: 

§ 140.10 Scope. 

This subpart applies to each person 

who is an applicant for or holder of a 

license issued under 10 CFR parts 50, 53 

or 54 to operate a nuclear reactor, or is 

the applicant for or holder of a 

combined license issued under 10 CFR 

parts 52, 53, or 54, except licenses held 

by persons found by the Commission to 

be Federal agencies or nonprofit 

educational institutions licensed to 

conduct educational activities. This 

subpart also applies to persons licensed 

to possess and use plutonium in a 

plutonium processing and fuel 

fabrication plant. 

■ 174. In § 140.11, revise paragraph (b) 

to read as follows: 

§ 140.11 Amounts of financial protection 

for certain reactors. 

* * * * * 

(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. 

■ 175. In § 140.12, revise paragraph (c) 

to read as follows: 

§ 140.12 Amount of financial protection 

required for other reactors. 

* * * * * 

(c) In any case where a person is 

authorized under 10 CFR parts 50, 52, 

53, or 54 to operate two or more nuclear 

reactors at the same location, the total 

financial protection required of the 

licensee for all such reactors is the 

highest amount which would otherwise 

be required for any one of those 

reactors; provided, that such financial 

protection covers all reactors at the 

location. 

* * * * * 

■ 176. Revise § 140.13 to read as 

follows: 

§ 140.13 Amount of financial protection 

required of certain holders of construction 

permits and combined licenses under 10 

CFR part 52. 

Each holder of a 10 CFR part 50 or 10 

CFR part 53 construction permit, or a 

holder of a combined license under 

parts 52 or 53 of this chapter before the 

date that the Commission had made the 

finding under §§ 52.103(g) or 53.1452(g) 

of this chapter, who also holds a license 

under part 70 of this chapter authorizing 

ownership, possession and storage only 

of special nuclear material at the site of 

the nuclear reactor for use as fuel in 

operation of the nuclear reactor after 

issuance of either an operating license 

under 10 CFR part 50 or 53, or a 

combined license under 10 CFR part 52 

or 53, shall, during the period before 

issuance of a license authorizing 

operation under 10 CFR part 50 or 53, 

or the period before the Commission 

makes the finding under § 52.103(g) or 

§ 53.1452(g) of this chapter, as 

applicable, have and maintain financial 

protection in the amount of $1,000,000. 

Proof of financial protection shall be 

filed with the Commission in the 

manner specified in § 140.15 before 

issuance of the license under part 70 of 

this chapter. 

■ 177. In § 140.20, revise paragraphs 

(a)(1)(i) and (ii) to read as follows: 

§ 140.20 Indemnity agreements and liens. 

(a) * * * 

(1)(i) The effective date of the license 

(issued under part 50 or part 53 of this 

chapter) authorizing the licensee to 

operate the nuclear reactor involved; or 

(ii) The date that the Commission 

makes the finding under §§ 52.103(g) or 

53.1452(g) of this chapter; or 

* * * * * 

PART 150—EXEMPTIONS AND 

CONTINUED REGULATORY 

AUTHORITY IN AGREEMENT STATES 

AND IN OFFSHORE WATERS UNDER 

SECTION 274 

■ 178. The authority citation for 10 CFR 

part 150 continues to read as follows: 

Authority: Atomic Energy Act of 1954, 

secs. 11, 53, 81, 83, 84, 122, 161, 181, 223, 

234, 274 (42 U.S.C. 2014, 2201, 2231, 2273, 

2282, 2021); Energy Reorganization Act of 

1974, sec. 201 (42 U.S.C. 5841); Nuclear 

Waste Policy Act of 1982, secs. 135, 141 (42 

U.S.C. 10155, 10161); 44 U.S.C. 3504 note. 

■ 179. In § 150.15, revise paragraphs 

(a)(7)(iii) and (a)(8) to read as follows: 

§ 150.15 Persons not exempt. 

(a) * * * 

(7) * * * 

(iii) Greater than Class C (GTCC) 

waste, as defined in part 72 of this 

chapter, in an ISFSI or an MRS licensed 

under part 72 of this chapter; the GTCC 

waste must originate in, or be used by, 

a facility licensed under parts 50, 52, or 

53 of this chapter. 

(8) Greater than Class C waste, as 

defined in part 72 of this chapter, that 

originates in, or is used by, a facility 

licensed under parts 50, 52, or 53 of this 

chapter and is licensed under part 30 

and/or part 70 of this chapter. 

* * * * * 

PART 170—FEES FOR FACILITIES, 

MATERIALS, IMPORT AND EXPORT 

LICENSES, AND OTHER 

REGULATORY SERVICES UNDER THE 

ATOMIC ENERGY ACT OF 1954, AS 

AMENDED 

■ 180. The authority citation for 10 CFR 

part 170 continues to read as follows: 

Authority: Atomic Energy Act of 1954, 

secs. 11, 161(w) (42 U.S.C. 2014, 2201(w)); 

Energy Reorganization Act of 1974, sec. 201 

(42 U.S.C. 5841); 42 U.S.C. 2215; 31 U.S.C. 

901, 902, 9701; 44 U.S.C. 3504 note. 

■ 181. In § 170.3, revise the definitions 

for ‘‘Manufacturing License,’’ ‘‘Part 55 

Reviews,’’ ‘‘Power reactor,’’ and 

‘‘Special projects’’ to read as follows: 

§ 170.3 Definitions. 

* * * * * 

Manufacturing license means a 

license under subpart F of part 52 of this 

chapter or subpart H of part 53 of this 

chapter to manufacture a nuclear power 

reactor(s) to be operated at sites not 

identified in the license application. 

* * * * * 

Part 55 Reviews as used in this part 

means those services provided by the 

Commission to administer 

requalification and replacement 

examinations and tests for reactor 


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Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules 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 §§ 50.21(b), 50.22, or 

part 53 of this chapter. 

* * * * * 

Special projects means specific 

services provided by the Commission 

for which fees are not otherwise 

specified in this chapter. This includes, 

but is not limited to, contested hearings 

on licensing actions directly related to 

U.S. Government national security 

initiatives (as determined by the NRC), 

topical report reviews, early site 

reviews, waste solidification activities, 

activities related to the tracking and 

monitoring of shipment of classified 

matter, services provided to certify 

licensee, vendor, or other private 

industry personnel as instructors for 10 

CFR part 55 or 53 reactor operators, 

reviews of financial assurance 

submittals that do not require a license 

amendment, reviews of responses to 

Confirmatory Action Letters, reviews of 

uranium recovery licensees’ land-use 

survey reports, and reviews of §§ 50.71 

or 53.1545 of this chapter Final Safety 

Analysis Reports. Special projects does 

not include activities otherwise exempt 

from fees under this part. It also does 

not include those contested hearings for 

which a fee exemption is granted in 

§ 170.11(a)(2), including those related to 

individual plant security modifications. 

* * * * * 

■ 182. In § 170.12, revise paragraph 

(d)(1)(v) to read as follows: 

§ 170.12 Payment of fees. 

* * * * * 

(d) * * * 

(1) * * * 

(v) 10 CFR 50.71 or 53.1545 final 

safety analysis reports; 

* * * * * 

§ 170.21 [Amended] 

■ 183. In § 170.21, in footnote 1 remove 

the phrase ‘‘(e.g., 10 CFR 50.12, 10 CFR 

73.5)’’ and add in its place the phrase 

‘‘(e.g., 10 CFR 50.12, 10 CFR 53.080, 10 

CFR 73.5)’’. 

■ 184. Revise § 170.41to read as follows: 

§ 170.41 Failure by an applicant or 

licensee to pay prescribed fees. 

If the Commission determines that an 

applicant or a licensee has failed to pay 

a prescribed fee required in this part, 

the Commission will not process any 

application and may suspend or revoke 

any license or approval issued to the 

applicant or licensee. The Commission 

may issue an order with respect to 

licensed activities that the Commission 

determines to be appropriate or 

necessary to carry out the provisions of 

this part, parts 30, 31, 32 through 35, 40, 

50, 53, 61, 70, 71, 72, 73, and 76 of this 

chapter, and of the Act. 

PART 171—ANNUAL FEES FOR 

REACTOR LICENSES AND FUEL 

CYCLE LICENSES AND MATERIALS 

LICENSES, INCLUDING HOLDERS OF 

CERTIFICATES OF COMPLIANCE, 

REGISTRATIONS, AND QUALITY 

ASSURANCE PROGRAM APPROVALS 

AND GOVERNMENT AGENCIES 

LICENSED BY THE NRC 

■ 185. The authority citation for 10 CFR 

part 171 continues to read as follows: 

Authority: Atomic Energy Act of 1954, 

secs. 11, 161(w), 223, 234 (42 U.S.C. 2014, 

2201(w), 2273, 2282); Energy Reorganization 

Act of 1974, sec. 201 (42 U.S.C. 5841); 42 

U.S.C. 2215; 44 U.S.C. 3504 note. 

■ 186. Revise § 171.3 to read as follows: 

§ 171.3 Scope. 

The regulations in this part apply to 

any person holding an operating license 

for a test reactor or research reactor 

issued under part 50 of this chapter, and 

to any person holding an operating 

license for a power reactor licensed 

under 10 CFR part 50 or 53, or a 

combined license issued under 10 CFR 

part 52 or 53, that has provided 

notification to the U.S. Nuclear 

Regulatory Commission (NRC) that the 

licensee has successfully completed 

power ascension testing. The 

regulations in this part also apply to any 

person holding a materials license as 

defined in this part, a Certificate of 

Compliance, a sealed source or device 

registration, a quality assurance program 

approval, and to a Government agency 

as defined in this part. Notwithstanding 

the other provisions in this section, the 

regulations in this part do not apply to 

uranium recovery and fuel facility 

licensees until after the Commission 

verifies through inspection that the 

facility has been constructed in 

accordance with the requirements of the 

license. 

■ 187. In § 171.5, revise the definitions 

for ‘‘Operating license,’’ and ‘‘Power 

reactor’’ to read as follows: 

§ 171.5 Definitions. 

* * * * * 

Operating license means having a 

license issued under §§ 50.57 or 53.1387 

of this chapter. It does not include 

licenses that only authorize possession 

of special nuclear material after the 

Commission has received a request from 

the licensee to amend its licensee to 

permanently withdraw its authority to 

operate or the Commission has 

permanently revoked such authority. 

* * * * * 

Power reactor means a nuclear reactor 

designed to produce electrical or heat 

energy and licensed by the Commission 

under the authority of section 103 or 

subsection 104b of the Atomic Energy 

Act of 1954, as amended, and under the 

provisions of §§ 50.21(b) or 50.22, or 

part 53 of this chapter. 

* * * * * 

■ 188. In § 171.15, revise paragraphs (a), 

(b)(2)(iii), (c)(1), and (d)(1) to read as 

follows: 

§ 171.15 Annual fees: Non-power 

production or utilization licenses, reactor 

licenses, and independent spent fuel 

storage licenses. 

(a) Each person holding an operating 

license for one or more non-power 

production or utilization facilities under 

10 CFR part 50 that has provided 

notification to the NRC of the successful 

completion of startup testing; each 

person holding an operating license for 

a power reactor licensed under 10 CFR 

part 50 or a combined license under 10 

CFR part 52, or an operating license or 

combined license for a commercial 

nuclear plant under 10 CFR part 53, that 

has provided notification to the NRC of 

the successful completion of power 

ascension testing; each person holding a 

10 CFR part 50 or 52, power reactor 

license, or a 10 CFR part 53 commercial 

nuclear plant license that is in 

decommissioning or possession only 

status, except those that have no spent 

fuel onsite; and each person holding a 

10 CFR part 72 license who does not 

hold a 10 CFR part 50, 52, or 53 license 

and provides notification under 

§ 72.80(g) of this chapter, shall pay the 

annual fee for each license held during 

the Federal fiscal year in which the fee 

is due. This paragraph (a) does not 

apply to test or research reactors 

exempted under § 171.11(b). 

(b) * * * 

(2) * * * 

(iii) Generic activities required largely 

for NRC to regulate power reactors (e.g., 

updating part 50, part 52, or part 53 of 

this chapter, operating the Incident 

Response Center, new reactor regulatory 

infrastructure). The base annual fee for 

operating power reactors does not 


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87128 Federal Register / Vol. 89, No. 211 / Thursday, October 31, 2024 / Proposed Rules 

include generic activities specifically 

related to reactor decommissioning. 

(c)(1) The FY 2022 annual fee for each 

power reactor holding a 10 CFR part 50 

operating license or combined license 

issued under 10 CFR part 52 or part 53 

that is in a decommissioning or 

possession-only status and has spent 

fuel onsite, and for each independent 

spent fuel storage 10 CFR part 72 

licensee who does not hold a 10 CFR 

part 50 or part 53 operating license, or 

a 10 CFR part 52 or part 53 combined 

license, is $227,000. 

* * * * * 

(d)(1) Each person holding an 

operating license for an SMR issued 

under 10 CFR part 50 or part 53, or a 

combined license issued under 10 CFR 

part 52 or part 53, that has provided 

notification to the NRC of the successful 

completion startup testing, shall pay the 

annual fee for all licenses held for an 

SMR site. The annual fee will be 

determined using the cumulative 

licensed thermal power rating of all 

SMR units and the bundled unit 

concept, during the fiscal year in which 

the fee is due. For a given site, the use 

of the bundled unit concept is 

independent of the number of SMR 

plants, the number of SMR licenses 

issued, or the sequencing of the SMR 

licenses that have been issued. 

* * * * * 

■ 189. In § 171.17, revise paragraphs (a) 

introductory text, (a)(1)(ii), and (a)(2) to 

read as follows: 

§ 171.17 Proration. 

(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 possessiononly license is received by the NRC, 

provided the licensee permanently 

ceased licensed activities during the 

specified period. The annual fee for 

non-power production or utilization 

facilities will be prorated based on the 

number of days remaining in the FY 

when the authorization to operate the 

facility has been permanently removed 

from the license during the FY. 

* * * * * 

Dated: October 7, 2024. 

For the Nuclear Regulatory Commission. 

Carrie Safford, 

Secretary of the Commission. 

[FR Doc. 2024–23434 Filed 10–23–24; 8:45 am] 

BILLING CODE 7590–01–P 


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