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

SUPPLEMENTARY INFORMATION:

Executive Summary

A. Need for the Regulatory Action

On January 14, 2019, the President signed the Nuclear Energy Innovation and Modernization Act (NEIMA) into law (Pub. L. 115-439). NEIMA section 103(a)(4) directs the NRC to “complete a rulemaking to establish a technology-inclusive, regulatory framework for optional use by commercial advanced nuclear reactor applicants for new reactor license applications.” NEIMA defines a “technology-inclusive regulatory framework” as one that is “developed using methods of evaluation that are flexible and practicable for application to a variety of reactor technologies, including, where appropriate, the use of risk-informed and performance-based techniques.” NEIMA, as further amended by the Accelerating Deployment of Versatile, Advanced Nuclear for Clean Energy Act of 2024 (ADVANCE Act), defines the term “advanced nuclear reactor” as “a nuclear fission reactor or fusion machine, including a prototype plant (as defined in sections 50.2 and 52.1 of title 10, Code of Federal Regulations (as in effect on the date of enactment of [NEIMA])), with significant improvements compared to commercial nuclear reactors under construction as of the date of enactment of [NEIMA].”

The NRC initially considered establishing the scope of proposed part 53, “Risk-Informed, Technology-Inclusive Regulatory Framework for Commercial Nuclear Plants,” of title 10 of the Code of Federal Regulations (10 CFR) as being for “advanced nuclear plants” consisting of one or more “advanced nuclear reactors” as defined in NEIMA. Based on public discussions on the use of the term, the NRC determined that the NEIMA definition, although broad, did not define “significant improvements” with enough specificity to implement in NRC regulations. Additionally, a number of stakeholders suggested that the descriptor, “advanced,” implied enhanced safety, while the NEIMA definition includes “significant improvements” in areas other than safety enhancements. In response to this feedback, and to be technology inclusive, the NRC determined that the broader term “commercial nuclear plant” would be preferable.

The current application and licensing requirements in 10 CFR part 50, “Domestic Licensing of Production and Utilization Facilities,” and 10 CFR part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” were primarily developed to address license requests concerning water-cooled reactors, and to address operational requirements for those types of reactors. This proposed rule responds to NEIMA by creating an alternative regulatory framework for licensing future commercial nuclear plants. The new alternative requirements and implementing guidance would adopt technology-inclusive approaches and use risk-informed and performance-based techniques to ensure an equivalent level of safety to that of operating commercial nuclear plants while providing flexibility for licensing and regulating a variety of technologies and designs for commercial nuclear reactors.

B. Major Provisions

Major provisions of this proposed rule, supported by accompanying guidance, include the following:

• A new alternative technology-inclusive, risk-informed, performance-based framework that includes requirements for licensing and regulating nuclear plants during the various stages of their life cycles.
• A new alternative technology-inclusive, risk-informed, and performance-based framework in10 CFR part 26, “Fitness for Duty Programs,” developed from existing requirements in subpart K, “FFD Programs for Construction,” of part 26.
• A new alternative technology-inclusive and performance-based security framework in10 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 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 provided the Commission's policy regarding the review of, and desired characteristics associated with, advanced reactors. The NRC updated this statement in the “Policy Statement on the Regulation of Advanced Reactors,” dated October 14, 2008 (Advanced Reactor Policy Statement).

The agency has undertaken many activities related to advanced reactors, including issuing an advance notice of proposed rulemaking titled, “Approaches to Risk-Informed and Performance-Based Requirements for Nuclear Power Reactors,” dated May 4, 2006 (71 FR 26267). These efforts were often done in parallel, and sometimes interwoven, with the NRC's efforts to improve risk-informed and performance-based approaches within the agency ( e.g., the Commission's policy statement, “Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities,” dated August 16, 1995 (PRA Policy Statement)).

In 2016, the NRC issued “NRC Vision and Strategy: Safely Achieving Effective and Efficient Non-Light-Water Mission Readiness” (Advanced Reactor Vision and Strategy Document), in response to increasing interest in advanced reactor designs. The NRC considered the Department of Energy's (DOE's) advanced reactor deployment goals in developing the Advanced Reactor Vision and Strategy Document. Since publication of the document, the NRC continues to manage its activities to support the DOE's deployment goals. The Advanced Reactor Vision and Strategy Document identified initiating and developing a new risk-informed and performance-based regulatory framework as a possible long-term goal. However, the NRC staff's initial efforts were focused on resolving policy issues and developing guidance for licensing non-LWR technologies under the existing regulatory frameworks (parts 50 and 52). The NRC staff issues annual Commission papers on the status and progress of the NRC staff's activities related to advanced reactors ( e.g., SECY-24-0020, “Advanced Reactor Program Status,” dated February 27, 2024). These Commission papers provide status updates for advanced reactor activities undertaken both prior to and after initiation of this rulemaking.

In 2017, the NRC staff prioritized activities to support the development of technology-inclusive, risk-informed, and performance-based licensing approaches that could be implemented under the existing regulatory framework in parts 50 and 52. One key element of these efforts was the Licensing Modernization Project (LMP), a cost-shared initiative led by nuclear utilities and supported by DOE. The LMP is a technology-inclusive, risk-informed, and performance-based methodology developed for non-LWR designs. The LMP provides a systematic and reproducible process for licensing-basis event (LBE) selection and evaluation; classification of structures, systems, and components (SSCs); and assessment of defense in depth. The LMP refined the DOE's Next Generation Nuclear Plant Program methodologies to reflect interactions with the NRC, to address feedback from industry, and to broaden the scope of the approach to ensure applicability to various non-LWR technologies. The LMP activities led to the publication and submittal of Nuclear Energy Institute (NEI) 18-04, Revision 1, “Risk-Informed Performance-Based Technology Inclusive Guidance for Non-Light Water Reactor Licensing Basis Development,” issued August 2019. The document indicates that controlling the frequencies and potential consequences of a wide spectrum of events is the primary focus of the LMP approach.

The NRC endorsed the principles and methodology in NEI 18-04, with clarifications, in RG 1.233, “Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors.” The NRC staff sought Commission approval of the use of LMP and NEI-18-04 in SECY-19-0117, “Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors,” dated December 2, 2019. In that paper, the staff described the relationship between the LMP and NEI-18-04 and previous relevant Commission decisions, including those described in SECY-93-092, “Issues Pertaining to the Advanced Reactor (PRISM, MHTGR, and PIUS) and CANDU 3 Designs and their Relationship to Current Regulatory Requirements,” dated April 8, 1993. The Commission approved the use of the LMP methodology and NEI-18-04 as a reasonable approach for establishing key parts of the licensing basis and content of applications for licenses, certifications, and approvals for non-LWRs in Staff Requirements Memorandum (SRM) SRM-SECY-19-0117, dated May 26, 2020. Although the LMP approach is technology- inclusive, the industry and NRC staff initially focused the LMP's applicability on non-LWRs, both for efficiency and to support near-term non-LWR applications under the existing regulatory framework, such as the Advanced Reactor Demonstration Projects supported by DOE.

As stated in the part 53 rulemaking plan, SECY-20-0032, the NRC staff developed part 53 by building upon recent and ongoing activities such as the LMP approach described in SECY-19-0117. Such an approach supports implementing the NEIMA requirement to use, where appropriate, risk-informed and performance-based techniques, and it also capitalizes on previous initiatives by the industry, DOE, and the NRC, including the LMP. This approach highlights the role of PRA in risk-informed and performance-based approaches to identifying enhanced safety margins that can be used to justify operational flexibilities. The proposed framework is largely based on the methodology described in SECY-19-0117 and includes a prominent role for PRA.

As discussed in section II.B, “Stakeholder Views on Part 53 Preliminary Proposed Rule Language,” of this document, the NRC conducted extensive public outreach on early versions of the proposed rule text. Early versions of the draft proposed rule included two alternative regulatory frameworks. One framework (called “Framework A”) offered a licensing approach centered largely on risk analysis and the other framework (called “Framework B”) largely replicated the existing licensing approach in parts 50 and 52 but modified it to be technology neutral. In its SRM to SECY-23-0021, “Proposed Rule: Risk-Informed, Technology-Inclusive Regulatory Framework for Advanced Reactors (RIN 3150-AK31),” the Commission disapproved the inclusion of Framework B in this proposed rule and directed the staff to provide them within one year an options paper for possible future use of the Framework B methodology.

B. Stakeholder Views on Part 53 Preliminary Proposed Rule Language

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

On November 6, 2020, the NRC published a notification in the Federal Register (85 FR 71002) describing plans for the periodic release of preliminary proposed rule language, meetings with stakeholders, and the ability of stakeholders to provide input during the development of this proposed rule. Sections of the preliminary proposed rule language were subsequently released, and the NRC held numerous public meetings to discuss the preliminary proposed rule language and obtain input from stakeholders. On December 10, 2021, the NRC published a second notification in the Federal Register (86 FR 70423) announcing that the development of the proposed rule and related interactions with stakeholders were being extended until August 31, 2022.

By the close of the public stakeholder interactions on August 31, 2022, the NRC staff had held 24 public meetings since September 2020. The NRC staff also met with the Advisory Committee on Reactor Safeguards (ACRS) in 16 public meetings during this period. By the close of the public engagement period on the preliminary proposed rule language, 126 letters were received on the preliminary proposed rule language. Of these 126 letters, 21 were from non-governmental organizations, 31 were from the public, one was from Congress, and the remaining 73 letters were from NRC licensees, the NEI, and other industry groups. In addition, the ACRS wrote four interim letter reports to the Chair on this rulemaking and issued its final letter report on November 22, 2022. The letters from stakeholders provided various points of view and suggestions for clarifications, additions, and deletions to the preliminary proposed rule language. Copies of these letters may be viewed and downloaded from the Federal rulemaking website https://www.regulations.gov, under docket number NRC-2019-0062. The inputs received were considered in the development of this proposed rule. However, as described during the various public interactions related to this rulemaking and in supporting documents, the NRC will not formally disposition the questions and suggestions related to the preliminary proposed rule language as it will for public comments received following the publication of this proposed rule.

III. Discussion

A. Objective and Applicability

The NRC is proposing to add a new, alternative part to its regulations that would set out a risk-informed, technology-inclusive framework for the licensing and regulation of commercial nuclear plants. This new approach would achieve the following: (1) continue to provide reasonable assurance of adequate protection of public health and safety and the common defense and security; (2) promote regulatory stability, predictability, and clarity; (3) reduce requests for exemptions from the current requirements in parts 50 and 52; (4) establish new requirements to address non-LWR technologies; (5) recognize technological advancements in reactor design; and (6) credit the possible response of some designs of commercial nuclear plants to postulated accidents, including slower transient response times and relatively small and slow release of fission products. This proposed rule would add 10 CFR part 53; subpart M, “Fitness for Duty Programs for Facilities Licensed Under 10 CFR Part 53,” to Part 26; § 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 technology-inclusive approach to the regulation of nuclear reactors in the advance notice of proposed rulemaking published in 2006. The NRC acknowledged in its “Report to Congress: Advanced Reactor Licensing,” issued August 2012, that while the safety philosophy inherent in the current regulations applies to all reactor technologies, the specific and prescriptive aspects of those regulations clearly focus on the current fleet of LWR facilities.

Congress similarly recognized the potential benefits of developing a regulatory infrastructure to support the development and commercialization of advanced nuclear reactors. Consequently, Congress passed NEIMA in late 2018, and the President signed it into law in January 2019. NEIMA directed the NRC to undertake a rulemaking to establish a technology-inclusive regulatory framework for optional use by applicants for new commercial advanced nuclear reactor licenses. In addition, on July 9, 2024, the President signed into law the Accelerating Deployment of Versatile, Advanced Nuclear for Clean Energy Act of 2024, also referred to as the ADVANCE Act. The NRC is evaluating its plans for implementing the ADVANCE Act, including how its regulations, as well as the proposed part 53 or future revisions to it, could be used to address provisions in the ADVANCE Act. The ADVANCE Act contains provisions on a variety of nuclear-related topics, such as micro reactors, nuclear reactor license application reviews, and nuclear fuel. In Section VI, “Specific Requests for Comments,” the NRC is requesting public input on how part 53 could be revised to better enable its potential use to implement the ADVANCE Act.

The requirements in part 53 would support a wide variety of potential commercial nuclear reactor technologies. As noted in this discussion, the current regulatory framework in parts 50 and 52 evolved in the context of the current operating reactor fleet dominated by LWRs and as a result includes provisions specific to LWR technologies. While the NRC can license other reactor technologies under the current framework by using existing regulatory flexibilities and the exemption process, there is significant interest in developing a regulatory framework that is flexible enough to accommodate multiple technologies and robust enough to ensure a level of safety equivalent to parts 50 and 52, consistent with the Commission's Advanced Reactor Policy Statement. The Commission reiterated its safety expectations for new reactors in the SRM for SECY-10-0121, “Modifying the Risk-Informed Regulatory Guidance for New Reactors,” dated March 2, 2011:

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

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

C. 10 CFR Part 53: Framework

This proposed rule consists of several major components, including a new part 53, to be added to 10 CFR chapter I, revisions for part 26, part 50, and part 73, and conforming changes throughout 10 CFR chapter I.

Part 53 is comprised of subparts A through M. These provisions are organized to provide high-level performance criteria and to specify requirements to demonstrate compliance with those performance criteria throughout major stages of the life cycle of commercial nuclear plants. This organization reflects a systems-engineering style approach to the design, licensing, operation, and ultimately decommissioning of future commercial nuclear plants. Organizing requirements in this manner also supports performance-based approaches. Required programs ( e.g., radiation protection) and monitoring ( e.g., technical specification (TS) surveillance) during the operations phase that are similar to those required by part 50 would complement the design and analysis requirements in subpart C. The performance-based approach proposed in part 53 also includes regulatory requirements that would allow applicants to use a flexible and graded approach to the performance of safety functions based on the role of a particular SSC, human action, or program in limiting the overall risks to the public below accepted standards through balanced measures to prevent and mitigate possible events.

Proposed subpart M of part 26 would be new and would be largely consistent with the objective-based fitness for duty (FFD) requirements in current subpart K, “FFD Programs for Construction,” of part 26 supplemented by select requirements from subparts A through I, N, and O of part 26. These requirements are designed to ensure program effectiveness, maintain protections afforded to individuals subject to the FFD program, and align with FFD program implementation by parts 50 and 52 licensees. The proposed requirements are not entirely equivalent because current subpart K of part 26 only applies during construction of the commercial nuclear plant, whereas proposed subpart M of part 26 would apply during construction, operation, and decommissioning. Furthermore, proposed subpart M of part 26 would allow the use of a variety of biological specimens for drug testing as well as innovative technologies for drug and alcohol screening and testing that are not described or allowed by the requirements in subparts A through K, N, and O of part 26, except under limited conditions.

Proposed revisions to part 73 would establish a new technology-inclusive consequence-based approach for a range of security areas, including physical security, cybersecurity, and access authorization (AA) for commercial nuclear reactors. The NRC used operating experience to include additional regulatory flexibility for a part 53 licensee's implementation of security requirements.

In addition, this proposed rule would make conforming changes throughout 10 CFR chapter I, by adding “and part 53” where appropriate to account for the addition of the proposed part 53.

IV. Part 53: Framework

Subpart A—General Provisions

Subpart A would provide the general provisions applicable to all applicants and licensees that would be established in part 53 for the issuance, amendment, and termination of licenses, permits, certifications, and approvals for commercial nuclear plants licensed under Section 103 of the Atomic Energy Act of 1954, as amended (the Act) and title II of the Energy Reorganization Act of 1974 (88 Stat. 1242). Subpart A would include purpose, scope, definitions, written communications, employee protections, completeness and accuracy of information, exemptions, standards for review, jurisdictional limits, consideration of attacks and destructive acts by enemies of the United States, and information collection requirements.

The requirements in subpart A would be largely equivalent to the general requirements in part 50 that are applicable to all part 50 applicants and licensees (specifically, §§ 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 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 non-safety-related but safety-significant (NSRSS) SSCs identified by the design and analysis requirements in subparts B and C to ensure the safety criteria are met.

Section 53.020 would also add definitions for terms related to event selection (LBEs, design-basis accidents (DBAs), anticipated event sequences, unlikely event sequences, and very unlikely event sequences); equipment classifications (SR, NSRSS, and non-safety-significant SSCs); performance metrics ( e.g., safety criteria and functional design criteria); and special treatment.

The regulation would define “ Safety criteria ” in terms of the plant-level performance-based metrics that would be provided in §§ 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 safety-significant SSCs ” would mean those SSCs that are not SR because they are not relied upon to perform any function necessary to demonstrate compliance with § 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-safety-significant SSCs ” would mean those SSCs that are not SR or NSRSS.

The terms “ Design-basis accidents, ” “ Anticipated event sequences, ” “ Unlikely event sequences, ” and “ Very unlikely event sequences ” would be defined to be different types of “ Licensing-basis events ” and would also be largely equivalent to the LMP's definitions of DBAs, anticipated operational occurrences (AOOs), design-basis events (DBEs), and beyond-design-basis events, respectively. The term “ 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-design-basis accidents terminology, these definitions seek to clarify the event categories in part 53. The sections of this preamble related to subparts B and C provide additional discussion of LBEs.

B. Other General Provisions

Section 53.040 would govern written communications and how applications and other required information must be submitted to the NRC. These requirements would be equivalent to those in § 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, “Technology-Inclusive Safety Requirements,” would provide technology-inclusive safety criteria that would serve as performance standards for the subsequent performance-based requirements used throughout part 53. Subsequent subparts would define how specific activities during various stages of the life cycle of a commercial nuclear plant contribute to satisfying these high-level performance standards. The performance standards in subpart B would also establish a means to determine appropriate regulatory controls for SSCs, human actions, and programs in the following subparts. For example, the classification of SR SSCs would be built upon the proposed safety criteria in § 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 performance-based requirements in part 53 would be sufficient, if met, for the NRC to make the findings required to grant an application for a utilization facility under Section 182 of the Act that the utilization of SNM will be in accord with the common defense and security and will provide adequate protection to the health and safety of the public. This construct would be similar to existing NRC regulations, which the Commission has said on many occasions do not specifically define “adequate protection.” However, compliance with NRC regulations may be presumed to assure adequate protection at a 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 decision-making. The methodology for developing and using proposed comprehensive risk metrics and associated risk performance objectives is defined by the proposed requirements for analyses in § 53.450. Therefore, the application must include a description of that methodology and, among other things, should explain the initial conditions, boundary conditions, and key assumptions used to develop and calculate the risk metrics. Screening tools and bounding or simplified methods may be used for any mode or hazard, provided that the applicant provides an acceptable technical basis. As with all risk-informed methodologies, treatment of uncertainties must be addressed.

The risk performance objectives established under this methodology are likely to involve assessing and averaging the risks over a period of time ( e.g., plant year) and would not constitute a real-time requirement that must be continuously demonstrated by the licensee. The use of a comprehensive risk metric or set of risk metrics and risk performance objectives that reflect an average risk to establish performance goals for SR and NSRSS SSCs is consistent with current practices that use other risk assessment techniques to address short-term plant configurations during plant maintenance activities.

It is worth noting that the evaluation of plant risks, as represented by a comparison of analysis results to acceptable risk performance objectives for comprehensive risk metrics, would be one of several performance standards used in subpart B. The proposed use of multiple performance standards, including deterministic criteria and defense-in-depth measures, reflects an integrated decision-making process similar to that described in RG 1.174, “An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,” Revision 3. The NRC's approval of using a comprehensive risk metric or set of metrics with associated risk performance objectives is not, by itself, an indicator of adequate protection. Rather, the comparison of comprehensive risk metrics to associated risk performance objectives that are acceptable to the NRC is part of a suite of regulatory requirements that, when considered holistically, form the basis for the NRC's decision-making. This is analogous to the approach used for plants licensed under part 50 and part 52, where no single regulatory requirement governs whether a plant is “safe enough.”

The RG 1.233, “Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors,” describes an example of an acceptable approach for identifying and analyzing LBEs under part 50 and part 52, including the use of the quantitative health objectives (QHOs) stated in the NRC's policy statement, “Safety Goals for Nuclear Power Plant Operation,” dated August 4, 1986 (51 FR 28044), as corrected and republished August 21, 1986 (51 FR 30028) (Safety Goals Policy Statement), as acceptable performance objectives for 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 then-current regulatory practices could lead to a more coherent and consistent regulation of nuclear power plants, a more predictable regulatory process, a better public understanding of the regulatory criteria that the NRC applies, and public confidence in the safety of operating plants. Accordingly, the Commission announced the safety goals with a focus on the risks to the public from nuclear power plant operation. Following the issuance of the Safety Goals Policy Statement, the NRC has used the comprehensive risk metrics and performance objectives provided in the safety goals within the criteria for many decisions involving safety judgments during the licensing and regulation of operating reactors and proposed nuclear reactor designs. Consistent with NUREG-0880, the proposed comprehensive risk metrics and associated risk performance objectives required under § 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, “Technology-Inclusive Identification of Licensing Events for Commercial Nuclear Plants.”

Section 53.250 would establish defense-in-depth requirements based on the longstanding philosophy of providing defense in depth to address uncertainties about the design, operation, and performance of commercial nuclear plants. For example, parts 50 and 52 address defense in depth through layered prescriptive technical requirements ( e.g., fuel performance, cladding integrity, reactor coolant system integrity, containment performance) for LWRs. In contrast, the flexibility afforded to applicants in how they propose to demonstrate compliance with the high-level safety criteria within part 53 would necessitate this specific requirement to ensure defense in depth is provided. The requirements in this section would state that no single engineered design feature, human action, or programmatic control, no matter how robust, should be exclusively relied upon to address LBEs other than DBAs. The phrase “engineered design feature” would not preclude the possible crediting of inherent characteristics within the design and analysis for commercial nuclear reactors. While defense in depth would only be assessed for LBEs other than DBAs, the need to ensure dedicated success paths for DBAs would contribute to the overall defense in depth for each commercial nuclear plant under part 53.

Section 53.260 would govern normal operations and would establish a level of safety based on current requirements in 10 CFR part 20, “Standards for Protection Against Radiation,” which limits doses to members of the public and dose rates in unrestricted areas.

Section 53.270 would provide for the protection of plant workers and would establish a level of safety based on current requirements in 10 CFR part 20 which limits occupational dose.

Subpart C—Design and Analysis Requirements

This subpart would provide requirements for the design of commercial nuclear plants and the supporting analyses, including the analyses of LBEs, to demonstrate that the performance standards in proposed 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 licensee-controlled programs that address LBEs other than DBAs. Specifically, SR SSCs, human actions, and programmatic controls needed to protect against DBAs are used to satisfy the safety criteria in § 53.210. Non-safety-related but safety-significant SSCs, human actions, and licensee-controlled programs that address LBEs other than DBAs generally contribute to the appropriate measures considering potential risks to public health and safety.

Section 53.400 would establish a requirement that design features be provided for each commercial nuclear plant to satisfy the safety criteria and fulfill safety functions from proposed subpart B during LBEs. Other sections in subpart C would, in turn, further address the necessary capabilities and reliabilities for SSCs by establishing functional design criteria, fulfilling design requirements, performing analyses of LBEs, performing other supporting analyses, and categorizing SSCs based on their roles in preventing or mitigating LBEs.

Section 53.410 would require that functional design criteria be defined for design features relied upon to demonstrate that the consequences from DBAs would be below the criteria in § 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) 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 long-term cooling. The requirements would be included to address the potential that some reactor designs may be able to achieve a stable end state for the purpose of event analyses but might need further actions to completely shut down and service the facility.
• § 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 technology-inclusive 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 in10 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 technology-inclusive 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 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 applicant-defined scope as part of determining the applicability of and conformance to consensus PRA standard supporting requirements consistent with the current state of practice. In addition, these determinations would include consideration of other aspects of the development of the PRA, such as PRA peer reviews. An NRC determination of the acceptability of a PRA includes but is not limited to assessing the initial and boundary conditions and key assumptions used in the analysis, treatment of uncertainties, and the use of screening tools and bounding or simplified methods for any mode or hazard, provided the use of those tools and methods is justified by an acceptable technical basis. In that regard, the consensus PRA standards would not be applied by the NRC as a strict checklist of requirements for part 53 PRA acceptability determinations.

The proposed § 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 risk-significant functions. These requirements would also provide confidence that the SSCs will perform under the service conditions and with the reliability credited in the analysis performed in accordance with § 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 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 risk-informed 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 Performance-Based Approach to Define the Site-Specific Earthquake Ground Motion,” would be an acceptable way to develop site-specific ground motion response spectra for SSCs under appendix S to part 50, which corresponds to SSCs that are categorized as the highest SDC (SDC-5) in ASCE/SEI 43-19.

The evaluation of seismic hazards under subpart D would need to be sufficient to inform a site-specific design ( e.g., a CP or custom COL) or confirm the use of a standard design for a commercial nuclear plant under § 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, 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 long-standing preference for siting reactors in areas of low population density would be maintained in part 53 by using the current language from part 100 in proposed § 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 well-defined command and control structure to manage construction activities. The requirements would generally reflect current requirements, with an emphasis on the quality assurance programs for complying with the requirements in appendix B to part 50. The proposed § 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 construction-related activities such as backfilling of excavated portions of the site appropriately address potential pre-construction activities such as the emplacement of retaining walls or drainage systems. Other requirements in these paragraphs would be equivalent to requirements in parts 50 and 52 with appropriate references to other parts for items such as possession of byproduct material or SNM, protecting operating units from construction activities for commercial nuclear plants with multiple reactor units, and having a redress plan in case LWA activities are terminated.

Section 53.610(c) would address inspection and acceptance activities by including requirements in part 53 equivalent to specific quality assurance criteria in appendix B to part 50 and inspections, tests, analyses, and acceptance criteria (ITAAC) in part 52 for COLs.

Section 53.620(a) would include proposed requirements covering the activities performed under an ML issued under part 53. Provisions related to MLs were first adopted by the NRC in 1973 through the addition of appendix M to part 50. The regulation supported the manufacture of a nuclear power reactor to be incorporated into a commercial nuclear plant under a CP and operated under an OL at a different location from the place of manufacture. 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 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 in-transit requirements for the physical protection of special nuclear material of moderate and low strategic significance,” on a case-by-case basis. Including appropriate security measures in the proposed part 53 regulations will provide additional openness and transparency for applicants applying for an ML who seek to load fuel into manufactured reactors at a manufacturing site.

Currently, § 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 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 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, ” “ Interaction-dependent-mitigation facility, ” “ Load following, ” “ Self-reliant-mitigation facility. ”

Sections 53.725 through 53.830 would be divided into four portions that would cover general operational requirements, operator and senior operator licensing requirements, generally licensed reactor operator (GLRO) requirements, and general training requirements for plant staff. The NRC intends to provide guidance addressing the review of operator staffing plans; the review of operator, senior operator, and GLRO examination programs; and the implementation of scalable HFE reviews. Licensees would be required to use GLROs upon demonstrating compliance with the criteria in § 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 technology-inclusive requirements for assessing the role of personnel in facility safety, applying human-system considerations within facility design, and incorporating operational approaches that are consistent with design-specific safety considerations. Most of these requirements would be adapted from portions of §§ 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 post-Three Mile Island requirements of § 50.34(f) in a technology-inclusive manner as detailed in the following:

• 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 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 technology-inclusive facility staffing will need to account for a potentially wide range of concepts of operations; for this reason, flexible and performance-based approaches for establishing required facility staffing are appropriate. However, once the appropriate facility staffing has been determined and approved by the NRC, such staffing must be maintained to ensure that the appropriately qualified individuals will be available when needed to support the safe operation of the facility. Therefore, the NRC is proposing under § 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 GLROs. While it is not anticipated that GLROs will have a role in the fulfillment of safety functions at self-reliant-mitigation facilities and, furthermore, that operators at such facilities would not be in a position by which to significantly influence radiological safety outcomes, the very nature of the § 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 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, plant-referenced simulators be used and would allow the use of alternative simulation facilities consisting of, for example, partial scope simulators or the plant itself, provided that all associated requirements can be demonstrated to be met using alternative approaches and methods. Additionally, in allowing for the possibility that an applicant or licensee might demonstrate compliance with training, examination, or experience requirements using the plant itself, the NRC is not allowing the initiation of transients on the actual plant. Consistent with this, aside from controlled reactivity manipulations that are conducted for the purposes of demonstrating compliance with experience requirements, actual plant components may not be operated for these purposes. Rather, the NRC perspective is that the use of the plant for training and examination purposes should be restricted to techniques such as walkthroughs, job performance measures, simulated tasks, use of augmented reality technology, and similar approaches that provide training and examination value while avoiding the operation of actual plant components.

There may be situations in which applicants for operator or senior operator licenses have previous training and experience that justifies waiving some, or all, of the initial examination requirements. The NRC is proposing to adapt the requirements of § 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 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 self-reliant-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-reliant-mitigation facilities.

In order to use GLROs in lieu of specifically licensed operators and senior operators, a OL/COL applicant would need to demonstrate that its proposed facility is a self-reliant-mitigation facility, i.e., that it will comply with the following requirements on an ongoing basis: maintaining GLRO qualifications for the performance of important functions and tasks; incorporating relevant programmatic controls into TS; administering the related programs for training, examination, and proficiency; and ensuring that the relevant provisions of parts 26 and 73 are met. Additionally, to provide for an accurate accounting of what individuals are licensed under the general license, facility licensees would be required to report the identities of all generally licensed reactor operators to the NRC on an annual basis. Furthermore, a facility licensee must ensure that the facility design and performance continue to meet the technological criteria to be classified as a self-reliant-mitigation facility ( i.e., the criteria of § 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 SAT-based training program for the initial training of GLROs that is adequate to ensure that they have the necessary knowledge, skills, and abilities to perform their duties. The NRC further proposes that such programs would be subject to NRC approval, oversight, and appropriate change control processes. The training program must ensure that 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 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 non-traditional personnel roles and responsibilities anticipated under proposed part 53; conversely, and for the same reasons, additional categories of plant personnel may need to be covered by part 53. The NRC also recognizes that the timeframe prescribed in § 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 defense-in-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 § 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 NRC-conducted FOF exercises are appropriate to assess the ability of a private security force of a licensed facility to defend against any applicable DBT. For the class of licensees that meet the criterion of § 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 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 NRC-conducted 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. 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 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).

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 well-established NRC practice. The rationale for binding the NRC staff and ACRS is similar to the rationale for a COL applicant referencing a standard design approval under part 52.

Proposed § 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 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 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 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 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 licensing-basis information obtained from the certified design would be divided into two categories. The most significant design information and the ITAAC would be certified by rule and designated as “certification information.” The remaining information, which makes up the majority of the design information approved as part of the DC, would not be certified by rule and is not considered “certification information.” Part 52 refers to these categories of information as Tier 1 and Tier 2 information, respectively, and refers to a change made to that information on a plant-specific basis as a departure. Under part 52, a departure from Tier 1 information requires an exemption and, 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 plant-specific departure from the information approved by the NRC as part of the DC rule but which is not certification information ( i.e., Tier 2 information under part 52) would be assessed using the process and criteria defined in § 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 risk-significant 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 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.

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 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 technology-inclusive, risk-informed, and performance-based approach for the application of drug and alcohol testing and fatigue management requirements for facilities licensed under part 53. The proposed requirements applicable to these applicants, licensees, and other entities would be commensurate with the radiological consequences presented by the applicants' facilities and the operation of these facilities. 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 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 security-related equipment or equipment that a risk-informed evaluation process has shown to be significant to public health and safety.

The second-tier requirements are based on the additional risk presented by nuclear reactor assembly, testing, fueling, and operation and the necessity for human actions in certain event sequences. The inclusion of the current part 26 requirements would align proposed part 53 FFD and AA program requirements with the current FFD and AA programs required for facilities licensed under parts 50 and 52. This approach would ensure effective and consistent AA and FFD program implementation across the commercial nuclear power industry, thereby ensuring uniform requirements for individuals who may perform roles and responsibilities for multiple facilities regardless of facility licensure.

Proposed § 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. 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 load-following; active and passive safety and security systems; innovative non-light-water heat transfer systems; and energy storage and hazardous chemical systems. The individuals at part 53 facilities may also be required to communicate to individuals both onsite and offsite, such as electrical load dispatchers, any conditions adverse to safety, security, or quality. Third, part 53 licensed facilities may be sited in geographically remote locations that may not have a physically available administrative or corporate support team to provide face-to-face oversight, engineering expertise, and maintenance support like that at NPUFs. This places a higher reliance on those individuals required at a part 53 facility being fit for duty and trustworthy and reliable because a replacement individual may not be readily available.

The NRC proposes to exclude drug and alcohol testing from the proposed § 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, fleet-level performance, if applicable, and industry performance. The licensee or other entity would be required to implement corrective actions if site FFD performance meets a licensee- or other entity-established threshold or to resolve a finding resulting from a qualitative review or audit in a manner that restores performance and corrects root causes, contributing causes, or both; (3) the requirements in proposed § 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 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. Small staffs at part 53 commercial nuclear plants may therefore be more susceptible to groupthink if they are working in an isolated environment where decision-making pressures may be high.

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. 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. Furthermore, the literature indicates that groups make riskier decisions than individuals acting alone due to the diffusion of responsibility among group members. ⁹ 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-for-duty 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 cross-references to existing part 26 requirements) for part 53 licensees and other entities. This should help licensees and other entities implement the requirements because it would enable easy cross-reference to similar requirements in other subparts that are being implemented by non-part 53 licensees and entities subject to part 26. Understanding how other licensees or other entities implement similar FFD requirements may facilitate the sharing of operating experience in program implementation.

The use of innovative technologies and a risk-informed performance-based framework parallels the considerations presented in the Advanced Reactor Policy Statement. As stated in the policy statement, “[S]implified systems should facilitate operator comprehension, reliable system function, and more straightforward engineering analysis.” Furthermore, these same attributes may reduce potential radiation exposures, help prevent the theft of nuclear materials, and use technology and design innovations. Should these components and systems be designed, implemented, and maintained to minimize reliance on human actions and leverage technology and innovation, then the robust and prescriptive FFD requirements in, for example, subparts B, “Program Elements,” and E of part 26 could be scaled to the part 53-licensed facility and its operation. This strategy would be implemented in the subpart M of part 26 framework.

Even though current subpart K of part 26, provides for FFD requirements commensurate with the radiological consequences presented by a nuclear power plant construction site, proposed subpart M of part 26 would not allow part 53 licensees and other entities to implement the requirements in subpart K. The principal reasons are that (without significant changes to subpart K that would be outside the scope of this rulemaking): (1) subpart K does not apply to holders of MLs who assemble or test a reactor; (2) subpart K only applies during construction, whereas subpart M would apply during construction, operation, and decommissioning through implementation of the insider mitigation program (IMP) required by § 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 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 remote-control station) for licensees or other entities licensed under part 53.

The framework of the current part 26 does not account for individuals who perform operating and maintenance duties at remote facilities. Although current § 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-in-depth regulatory framework that helps provide that individuals who have unescorted access are fit for duty, trustworthy, and reliable. For example, the NRC is proposing amendments to part 73 to require a part 53 licensee to subject individuals to a series of reviews to help determine whether those individuals are trustworthy and reliable before granting them unescorted access to the facility's protected area.

The NRC would amend § 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 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 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 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 human-related failure causing a condition adverse to safety and security is dependent on the duties and responsibilities of the individual and the defense-in-depth designed to prevent or mitigate an adverse consequence. The PMRP would account for this by requiring the review of FFD performance data, in part, by work category, C/V, and individuals employed by the licensee who are not a C/V as defined in § 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 entity-designated individuals. The other performance measures would include occurrence of FFD policy violations evaluated by licensee employee, C/V, and labor category, and occurrence of individuals with potentially disqualifying information or who possessed an FFD prohibited item.

Proposed § 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 performance-based regulatory strategy in which the licensee or other entity must initially establish a level of performance that is representative of other facilities in the licensee's fleet of facilities subject to part 26, if applicable, and the FFD performance of comparable facilities subject to part 26.

Proposed § 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 protection ( i.e., security personnel), or safety ( i.e., NRC-licensed operators and individuals who perform or direct the performance of activities that a risk-informed evaluation process has shown to be significant to public health and safety), a single FFD policy violation could be a significant occurrence and warrant corrective actions. Based on licensee-submitted FFD-related reports under §§ 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-to-year comparison of FFD performance data to help provide assurance that an adverse trend in FFD program performance would be identified if occurring. This proposed requirement was developed from the annual FFD performance data reporting requirements in §§ 26.417(b)(2) and 26.717. In particular, the proposed year-to-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 NRC-licensed operator or maintenance professional had operated, surveilled, or maintained safety-significant SSCs and was determined to have been impaired by behavioral observation or potentially under the influence of a narcotic as determined by an alcohol or drug test or screening result. Immediate corrective actions could include, but would not be limited to, a licensee or other entity assessment of the duties and responsibilities recently performed by the individual. Operating experience within the LWR operating reactor community demonstrates few FFD policy violations per year per site have been caused by individuals who perform or direct the performance of 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. 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 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-in-depth measures proposed in subpart M of part 26 like the PMRP, change control, and audits; and other requirements, such as those for AA, physical protection, and licensed operators. The adequacy of these measures in satisfying the performance objectives is supported by operating experience, which demonstrates margin between an FFD-related occurrence and a condition adverse to safety or security, as illustrated by for-cause, post-event, and random testing data. A facility that satisfies the criterion in proposed § 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.

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 security-related equipment or equipment that a risk-informed evaluation process has shown to be significant to public health and safety. These entities must establish, implement, and maintain an FFD program that implements all the requirements in § 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 NRC-licensed 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 follow-up 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 entity-designated 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, 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.

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 § 73.100(b)(9). 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 security-related SSCs” in current § 26.403(b)(3), the NRC would use “on the NRC-licensed 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 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 HHS-certified 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 security-related 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 “pre-assignment,” 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 pre-access testing must be performed from 60 days to 30 days or less, the 14-day proposal is based on three lessons learned from operating experience.

First, the 14-day period would be a large enough window of time to collect the specimen and evaluate test results because licensees or other entities typically receive laboratory test results within 5 business days of laboratory receipt of the biological specimen. At the same time, the 14-day period would be small enough to help ensure that the test results are representative of the individual's forensic toxicology before being granted authorization.

Second, the 14-day window would enable the licensee or other entity to conduct an unannounced pre-access drug and alcohol screening using a hair specimen or a POCTA. This would help prevent an individual from attempting to subvert the drug and alcohol test by temporarily abstaining from drug or alcohol abuse or adulterating or substituting their specimen to obtain a non-positive test result.

Third, the NRC does not expect licensees and other entities licensed under part 53 to have the large and periodic influxes of individuals (either licensee employees or C/Vs) that LWRs have to support facility operation, maintenance, engineering design changes, or nuclear refueling. Therefore, these licensees or other entities would not be periodically challenged to in-take a large workforce within the proposed 14-day pre-access testing window.

Proposed § 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 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 post-event testing) and offsite (for pre-access, post-event, and follow-up testing) collection facilities for drug and alcohol testing and may have to afford reasonable accommodation to certain individuals, which would add complexity in the licensee's or other entity's procedurally determined time period in which an individual must report to the collection facility.

Another difference from § 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 post-event 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 post-event 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 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), MRO-directed tests, and the testing of a split specimen. Further, HHS-certified laboratory test results using urine or oral fluid would be required for the issuance of an FFD policy violation and part 26-required sanction.

All drug testing would need to be performed at an HHS-certified laboratory to help ensure FFD program effectiveness and to protect the donor from a false positive test result and an unwarranted FFD policy violation. The donor would be protected because laboratory procedures for specimen accessioning, testing, custody and control, and evaluation of test results and the training and qualification of laboratory personnel are evaluated by HHS as part of the NLCP. This provides assurance that the drug testing results are accurate and attributed to the donor. Urine, oral fluid, and hair specimens may also be screened and tested for drugs and alcohol as described in § 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 HHS-certified, 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 HHS-certified laboratory.

Proposed § 26.607(c)(4) would require the licensee or other entity to contract with a primary and backup HHS-certified laboratory. This provision would help ensure that specimens are processed and tested to maintain FFD program effectiveness should the primary laboratory be unable to perform specimen testing. This would help maintain protections afforded to individuals subject to the FFD program ( e.g., should the donor or MRO request testing of the split specimen, a different laboratory could be used). This requirement also would state that the primary and backup laboratories must have a different certifying scientist. Having a back-up HHS-certified laboratory and a different certifying scientist would benefit the program and donor because the drug testing instruments, technicians, and certifying scientist would be independent of the 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 NLCP-identified weaknesses ( e.g., custody and control, accessioning, instrumentation, procedures, training, supervision, review of test results, and resolution of previously identified corrective actions) that may impact the effectiveness of FFD programs.

Proposed § 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 cost-effective 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 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 laboratory-confirmed positive test result, an FFD policy violation may not be administered. This laboratory information must be treated as potentially disqualifying FFD information, unless the individual subverts the screening process, in which case a permanent denial of authorization must be issued under proposed § 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 HHS-certified laboratory must be used to analyze the hair specimen, a forensic toxicologist must review the licensee's or other entity's hair screening process, the test kit must be cleared by the FDA, and hair screening must be conducted in accordance with the HHS Guidelines. The forensic toxicologist review would be necessary if the panel of drug or drug metabolites to be tested and their cutoffs are not established by HHS or the NRC for hair.

Proposed § 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 NRC-licensed facility to perform or direct those duties and responsibilities or maintain those types of access making them subject to the FFD program are fit for duty and deterred from arriving onsite in a physiological condition that may be adverse to safety and security. Additionally, screening could be conducted when an individual exits the NRC-licensed facility to provide assurance that substance abuse had not occurred on the site (see § 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 entity-designated MRO and not one designated by a third party, such as an MRO 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 medical- or clinical-based training session on a triennial basis. This proposal was developed from Section 13.1 of the HHS Guidelines for urine and oral fluid with two substantial differences: the HHS Guidelines state that “requalification training,” including an exam, must be conducted “at least every 5 years from initial certification,” whereas the proposed § 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 entity-designated MRO to determine and approve the use of oral fluid or urine as an alternative biological specimen when the donor cannot provide a requested specimen for testing. This proposed requirement is equivalent to § 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.

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 licensee- or other entity-designated and -trained individual. For subpart M of part 26, this would include onsite specimen collections, except a collection by a portal area screening instrument in § 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-duty-related information would include drug and alcohol testing results (not quantitative testing values), issuance of any sanctions, FFD-determinations regarding trustworthiness and reliability, testing programs, treatment, and other remedial or corrective action.

Proposed § 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 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 entity-designated representative(s). The intent of the BOP requirement is not to require that all individuals be observed at all times by others; NRC-licensed operators, maintenance professionals, security officers, and others routinely perform solo operations periodically throughout the day. However, individuals must be subject to observation while they are performing or directing the performance of duties and responsibilities or maintaining the types of access making them subject to the FFD program. Observing behavior only at the beginning of a work shift is not sufficient to ascertain whether an individual is fit for duty, trustworthy, and reliable. Controlled substances may have a delayed effect between use ( e.g., ingestion) and the onset of physiological or psychological effects, and fatigue accumulates with time. Behavior must be continually observed throughout the work shift to detect any changes from baseline human performance characteristics, including mental or physical health and mannerisms, or any activities that may indicate that the individual is not trustworthy and reliable.

Proposed § 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 security-related 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, in-person, 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 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 NRC-licensed facility for certain violations of the FFD policy within the protected area of a commercial nuclear plant and by an individual or individuals who are the operators of the conveyance to transport or use formula quantities of strategic SNM. Equivalent to § 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 HHS-certified 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 HHS-certified 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 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 NRC-licensed facility subject to subpart C of part 26, complete their NRC-required sanctions and licensee-administered or -directed drug and/or alcohol abuse treatment plans before the restoration of authorization by a licensee or other entity. Information sharing may also enhance FFD program effectiveness because FFD-related lessons learned from, for example, substance testing, subversion attempts, and laboratory and MRO performance must be shared when requested.

Proposed § 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. Based on this study, if electronic means were to be used to conduct a face-to-face assessment, an in-person element would still be integral to the assessment process. However, under certain circumstances, face-to-face determinations and assessments conducted as part of an FFD program for an entity licensed under part 53 ( i.e., those determinations and assessments performed in accordance with § 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 performance-based, technology-inclusive, risk-informed, 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 pre-determined, 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 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 sabotage-related events are not bounded by the consequences of the unlikely and very unlikely event sequences analyzed under subpart C. Accordingly, the NRC plans to develop guidance for part 53 applicants and licensees choosing to comply with the alternative emergency preparedness requirements in § 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 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-in-depth 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 defense-in-depth philosophy applies to the design of a physical protection program, which integrates engineered controls and administrative controls, to provide protection against the DBT for radiological sabotage.

Section 73.100(b)(3) would require the physical protection program to be designed and implemented to achieve and maintain the reliability and availability of SSCs required for demonstrating compliance with specified performance requirements. These physical protection performance requirements were informed by § 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. 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 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 non-LWR and SMRs, target sets would be defined in DG-5071, “Target Set Identification and Development for Nuclear Power Plants,” as the minimum combination of equipment, operator actions, and/or structures that, if all are prevented from performing their intended safety function or prevented from being accomplished, barring extraordinary actions by plant operations, would likely result in a significant release of radionuclides from any source ( e.g., a release to the environment exceeding that analyzed in the DBA licensing basis).

Section 73.100(b)(5) would require that each licensee perform a site-specific analysis for the purpose of identifying and analyzing site-specific conditions that affect the design of the onsite physical protection program.

Section 73.100(b)(6) would require licensees to implement a performance evaluation program, which would ensure that a licensee will periodically test and evaluate the effectiveness of the physical protection program to protect against the DBT. This program would ensure that licensees are able to demonstrate that the physical protection program satisfies the response requirements of § 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 self-assessments of physical protection program functions to ensure that the capability to detect, assess, interdict, and neutralize the DBT of radiological sabotage is maintained. Section 73.100(f)(3) would require an audit of the effectiveness of the physical protection program; security plans; implementing procedures; cybersecurity programs; management of the safety/security interface activities; the testing, maintenance, and calibration program; and response commitments by local, State, and Federal law enforcement authorities. Section 73.100(f)(4) would require that results and recommendations, management findings, and any actions taken be documented and maintained to be available for inspection by the NRC. These reviews are independent of the ongoing performance evaluations described in § 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 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 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 five-year reinvestigation periodicity for personnel at an operating commercial nuclear plant. In addition, no credit check re-evaluation would be required for these individuals.

The continued need to maintain unescorted access would be evaluated on an annual basis by the reviewing official. Guidance in DG-5074, “Access Authorization Program for Commercial Nuclear Plants,” would specify that this evaluation should be based on a compilation of personnel interactions as described in the licensee's or applicant's policy and procedures for behavioral observation and the maintenance of an approved AA list.

Section 73.120(c)(5) would require licensees and applicants to determine when a person no longer requires the need for unescorted access or no longer satisfies the AA requirement found within this section. Guidance in DG-5074 would further explain that licensees have the flexibility to terminate unescorted access to specific areas of the site if individuals lack the continued need for that access to perform their duties and responsibilities.

Section 73.120(c)(6) would be consistent with the purpose of § 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 licensee or other entity performs a site-specific 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 NRC-licensed operator regulatory framework of part 53. Please provide your considerations and rationale for your recommendation.

Should the NRC also consider making a conforming change to the proposed § 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 industry-developed guidance. The NRC may engage with the industry regarding a potential update to industry guidance document NEI 06-11 or the development of new, separate industry-developed guidance specific to facilities licensed under part 53. The NRC would then review the new or revised industry-developed guidance within the NRC's RG process, which includes opportunities for public participation. New or revised industry-developed guidance could incorporate DG-5078 or propose alternatives for the NRC to consider.
• Option 4: Develop a comprehensive revision of the existing RG. The NRC may develop a more comprehensive revision of RG 5.73 that would explicitly detail all NRC positions reflected in the existing RG (including those endorsed positions currently contained in NEI 06-11, Revision 1), along with the guidance of DG-5078. Such a revision would thereby be a “stand-alone” document, without reference to or explicit endorsement of separate, industry-developed guidance.

The NRC is seeking stakeholder input regarding which of the four options listed above would be optimal (or whether there are other options that the NRC should consider). Please provide your considerations and rationale for your recommendation.

Part 53—Overall Organization

Part 53 is structured as one framework with subparts providing technical, licensing, and administrative requirements for the various stages of the life cycle of a commercial nuclear plant. The organization of part 53 in this manner puts a complete set of requirements for each stage of the life cycle in a separate subpart with additional subparts for licensing and administrative requirements.

The NRC is seeking comment on the proposed organization of the requirements in part 53 and possible improvements to how specific requirements ( e.g., examples of which specific sections) could be consolidated or otherwise reorganized to make the rule clearer or more concise.

There are numerous references in proposed part 53 to other NRC regulations. Examples of such references include those in proposed § 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 extent practicable. Associated risk performance objectives are preestablished values indicative of the comprehensive risk metrics that are used during risk-informed decision-making to gauge plant safety. Specifically, comprehensive risk metrics and associated risk performance objectives would provide one element of the safety criteria for LBEs other than DBAs in the proposed § 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 decision-making process is similar to that used in RG 1.174, “An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,” Revision 3.

The NRC is seeking comment on the use of comprehensive risk metrics and associated risk performance objectives in part 53 as one of several performance standards. The IEFR and ILCFR and the QHOs represent comprehensive risk metrics and associated risk performance objectives that the NRC has used for decades in a variety of capacities. What other performance standards could be used to address the comprehensive risks posed by proposed commercial nuclear plants? Please provide your considerations and rationale for your recommendation.

If an applicant proposes a novel approach to comprehensive plant risk and the NRC approves the approach, should the resulting NRC-approved comprehensive plant risk metrics and associated risk performance objectives be codified or otherwise memorialized over time and, if so, how?

Part 53, Subpart B—Defense in Depth

Proposed § 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 micro-reactor designs. However, as written, the proposed provisions are not limited to any size or type of reactor.

The NRC is seeking comment on whether the proposed regulations are sufficient to govern various scenarios for the possible manufacturing and deployment of manufactured reactors.

If a comment indicates that the proposed regulations are not sufficient, please describe the reasons why, including, if applicable, any plausible scenario for which the commenter believes the proposed regulations are not sufficient.

2. The proposed regulations in subpart H allow holders of or applicants for a COL to reference an ML but do not include such a provision for the holder of or applicant for a CP or OL. This proposed change from the current relationship between subparts in part 52 and the part 50 licensing process was made to simplify the provisions in the proposed part 53 for licensing and deploying manufactured reactors.

The NRC seeks comment on whether part 53 should include provisions for an applicant for or a holder of a CP or an OL to reference an ML and, if so, how this should be done.

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 ( 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.

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 security-related events for an advanced reactor that addresses security through design and engineered safety features when it harmonizes this rulemaking with the EP for SMR/ONT rule. In the EP for SMR/ONT rule, the NRC established an alternative performance-based and risk-informed approach for emergency planning, including determining the need for and size of an emergency planning zone (EPZ) to support predetermined, prompt protective actions. The NRC has incorporated the relevant rule language from the EP for SMR/ONT rule into this proposed rule and is seeking stakeholder feedback as to whether additional rule language changes or additional guidance would be beneficial.

In light of the Commission direction and the above considerations, the NRC is assessing how best to address the treatment of security-related events in emergency planning, including in the determination of EPZ size, for reactors licensed under part 53. Part 53 is introducing an alternative approach to meeting security regulations that should be taken into consideration under § 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;
• 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 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 PRA-related information, the NRC is seeking comment on the appropriate control of that information and on the routine submittal of updates to the NRC. Please provide your considerations and rationale for your answer.

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

Proposed § 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? 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.

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 Micro-Reactors,” of the ADVANCE Act specifically mention the technology-inclusive regulatory framework to be established under section 103(a)(4) of NEIMA as a potential vehicle to be considered for the report to Congress required under section 203 and a potential vehicle to implement strategies and guidance for the licensing and regulation of micro-reactors required under section 208. This proposed rulemaking is, in part, how the NRC is implementing section 103(a)(4) of NEIMA.

The NRC is seeking comment on how part 53 could be revised to better enable its potential use to implement the ADVANCE Act. Specifically, Section 208 of the ADVANCE Act requires the NRC to develop and implement “risk-informed and performance-based strategies and guidance” in several areas for the licensing and regulation of micro-reactors, including with respect to “licensing mobile deployment.” The ADVANCE Act requires the NRC to consider “the unique characteristics of micro-reactors,” including physical size, design simplicity, and source term; opportunities to incorporate specific improvements related to streamlining the review process; and other policy and licensing issues. With regard to implementation, the ADVANCE Act provides the NRC with three options. The NRC may implement the developed strategies and guidance, as appropriate, via (1) the existing regulatory framework, (2) the Part 53 rulemaking, or (3) a pending or new rulemaking. Given the language included in Section 208, the NRC is seeking comment on how part 53 could be revised to better address the ADVANCE Act's requirements related to strategies and guidance for micro-reactors.

VII. Section-by-Section Analysis

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

§ 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.

§ 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, full-power 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 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.

§ 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.

§ 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-Light-Water 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.

§ 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, performance-based framework for the issuance, amendment, renewal, and termination of licenses, permits, certifications, and approvals for commercial nuclear plants licensed under section 103 of the Atomic Energy Act of 1954, as amended.

Subpart A—General Provisions

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

§ 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 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 technology-inclusive 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 Design-Basis 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 Licensing-Basis 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 performance-based, 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 Licensing-Basis 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 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 site-specific assessments, including an evaluation of geological and seismic siting factors, to identify and characterize the external hazard level for a range of natural and constructed hazards.

§ 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 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-dependent-mitigation 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 Self-Reliant-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 self-reliant 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 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 site-specific 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, post-shutdown 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 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 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.

§ 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 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.

§ 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 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 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 seventy-five 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.

§ 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 In-Transit 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 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 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 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.

The form number if applicable: NRC Forms 361S, 366, 366A, 366B, 893, and 894.

How often the collection is required or requested: Once, on occasion, every 30 days, biannually, annually, biennially, every four years, every five years, every ten years.

Who will be required or asked to respond: Part 53 commercial nuclear plant licensees and license applicants for commercial nuclear plants to be licensed under part 53.

An estimate of the number of annual responses: 15 (2 responses for Part 26, 11 responses for Part 53, 2 responses for Part 50 and 0 responses for Part 73 and NRC Forms 361S, 366, 366A, 366B, 893, and 894)

The estimated number of annual respondents: 2 (2 respondents for Part 26, 2 respondents for Part 53, 2 respondents for Part 50 and 0 respondents for Part 73 and NRC Forms 361S, 366, 366A, 366B, 893, and 894)

An estimate of the total number of hours needed annually to comply with the information collection requirement or request: 230,244 hours. (656 hours for Part 26, 220,801 hours for Part 53, 8,767 hours for Part 50 and 0 hours for Part 73 and NRC Forms 361S, 366, 366A, 366B, 893, and 894)

Abstract: The NRC is proposing to establish an optional technology-inclusive regulatory framework for use by applicants for new commercial nuclear plant designs. The regulatory requirements developed in this rulemaking would use methods of evaluation, including risk-informed and performance-based methods, that are flexible and practicable for application to a variety of new reactor technologies. The NRC's goals in amending these regulations are to continue to provide reasonable assurance of adequate protection of public health and safety and the common defense and security at reactor sites at which new nuclear reactor designs are deployed to at least the same degree of protection as required for current-generation LWRs; protect health and minimize danger to life or property to at least the same degree of protection as required for current-generation LWRs; provide greater operational flexibilities where supported by enhanced margins of safety that may be provided in new nuclear designs; and promote regulatory stability, predictability, and clarity.

The proposed rule covers diverse topics, which result in recordkeeping and reporting requirements related to contents of applications, plant design and analysis, siting, construction and manufacturing, licensing-basis information, facility operations, programs, staffing, FFD, physical security, cyber-security, AA, decommissioning, and quality assurance.

In addition to the new information collections in the proposed regulations, part 53 would result in new collections via NRC Forms 361S, 366, 366A, 366B, 893, and 894. NRC Forms 366, 366A, and 366B would be modified to include part 53 reportable events covering an equivalent scope as the requirements in 10 CFR 50.73, but without LWR-specific terminology to ensure technology inclusiveness. The proposed rule also would require part 53 licensees to use NRC Forms 893 and 894 to report on positive drug and alcohol test results (NRC Form 893) and annual fitness-for-duty program performance (NRC Form 894). Finally, a new version of NRC Form 361 (NRC Form 361S) would be created for use by part 53 licensees, covering an equivalent scope as the requirements in 10 CFR 50.72, but without LWR-specific terminology to ensure technology inclusiveness.

The NRC is seeking public comment on the potential impact of the information collections contained in this proposed rule and on the following issues:

1. Is the proposed information collection necessary for the proper performance of the functions of the NRC, including whether the information will have practical utility? Please explain your response.

2. Is the estimate of the burden of the proposed information collection accurate? Please explain your response.

3. Is there a way to enhance the quality, utility, and clarity of the information to be collected? Please explain your response.

4. How can the burden of the proposed information collection on respondents be minimized, including the use of automated collection techniques or other forms of information technology? Please explain your response.

The OMB clearance documents and proposed rule is available as indicated under the “Availability of Documents” section in this document or may be viewed free of charge by contacting the NRC's PDR reference staff at 1-800-397-4209, at 301-415-4737, or by email to 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 built upon recent and ongoing activities such as those described in SECY-19-0117. Because a number of those activities are ongoing to support new reactor applications under the existing regulatory framework of 10 CFR parts 50 and 52, the NRC staff identified in its response to SRM-SECY-20-0032 that the timing of guidance document development to support the part 53 rulemaking was a key risk and uncertainty to publishing the final part 53 rule. To mitigate this risk, the NRC engaged external stakeholders to ensure a common prioritization of the development of these guidance documents and to work diligently on those that would be needed to support this rulemaking, forthcoming applications, or broader efforts such as the Advanced Reactor Demonstration Program being sponsored by the DOE. The NRC also recognizes that guidance development to support part 53 and advanced reactors will continue as the industry and NRC learn lessons from licensing reviews and operating experience. Therefore, the NRC categorized guidance supporting the part 53 rulemaking into three categories: (1) guidance issued or under development to support applications under the existing regulatory framework; (2) implementing guidance for part 53-specific proposed rule language; and (3) future guidance activities that would need to be completed after the part 53 proposed rule is published for public comment.

(1) Hundreds of guidance documents exist for the current fleet of operating reactors. While some of the guidance is specific to LWR technologies, other guidance is technology inclusive in nature and should be considered, as appropriate, in the development of all licensing applications and NRC reviews. In addition, the NRC has undertaken efforts to incorporate or reference the most relevant guidance in its efforts to develop additional guidance for future advanced reactors. The NRC has issued the following guidance to support licensing reviews of advanced reactors under the existing regulatory framework that will continue to inform applicant development and NRC reviews under parts 50 and 52. Conforming changes to these guidance documents would be needed to ensure they are applicable under part 53. The NRC will issue revisions or part 53-related companions to these guidance documents for public comment after the publication of this proposed rule and then finalize and issue the guidance documents with or after the final part 53 rule.

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

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

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

(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 Under10 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 Under10 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 Under10 CFR part 53”

This DG describes proposed methods that the NRC staff considers acceptable for addressing certain aspects of FFD programs that would be established at commercial nuclear facilities licensed under part 53. This guidance, in conjunction with the existing RG 5.73, “Fatigue Management for Nuclear Plant Personnel,” would provide comprehensive guidance regarding acceptable methods for the development and implementation of licensee fatigue-management programs.

The NRC is issuing for public comment the following draft ISG documents for the implementation of NRC staff review of applications under the proposed requirements in this rulemaking:

• DRO-ISG-2023-01, “Operator Licensing Programs”

This draft ISG provides guidance for the review of tailored operator licensing programs that are submitted for review consistent with the technical requirements of proposed § 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 in10 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 staffing-related considerations, such as provisions for engineering expertise.

• DRO-ISG-2023-03, “Development of Scalable Human Factors Engineering Review Plans”

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/public-involve/public-meetings/index.cfm.

XIX. Availability of Documents

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