Modernizing Reactor Licensing, Safety Oversight, and Siting Practices

NUCLEAR REGULATORY COMMISSION 10 CFR Parts 2, 50, 51, 52, 53, 54, 71, and 100 [NRC-2025-0975] RIN 3150-AL44 Modernizing Reactor Licensing, Safety Oversight, and Siting Practices AGENCY:

Nuclear Regulatory Commission.

ACTION:

Proposed rule and guidance; request for comment.

SUMMARY:

Consistent with Executive Order 14300, “Ordering the Reform of the Nuclear Regulatory Commission,” the U.S. Nuclear Regulatory Commission (NRC) is conducting a review and wholesale revision of its regulations. This proposed rule aims to modernize reactor licensing, safety oversight, and siting practices addressing sections 5(f), 5(h), and 5(i) of Executive Order 14300, and additional items that contribute to adding additional generation to the electrical grid. Additionally, as part of the NRC's overarching review of all of its regulations, the agency identified a number of further changes to the NRC's regulations that will improve the efficiency and efficacy of its licensing process that are also included in this rulemaking.

DATES:

Comments must be submitted electronically using https://www.regulations.gov by 11:59 p.m. eastern time on August 31, 2026. Comments received after this date will be considered if it is practical to do so, but the Commission is able to ensure consideration of only comments received before this date.

ADDRESSES:

Submit your comments, identified by Docket ID NRC-2025-0975, at https://www.regulations.gov. If your material cannot be submitted using https://www.regulations.gov, call or email the individuals listed in the FOR FURTHER INFORMATION CONTACT section of this document for alternate instructions.

Do not include any personally identifiable information (such as name, address, or other contact information) or confidential business information that you do not want publicly disclosed. All comments are public records; they are publicly displayed exactly as received, and will not be deleted, modified, or redacted. Comments may be submitted anonymously.

Follow the search instructions on https://www.regulations.gov to view public comments.

You can read a plain language description of this proposed rule at https://www.regulations.gov/docket/NRC-2025-0975. For additional direction on obtaining information and submitting comments, see “Obtaining Information and Submitting Comments” in the SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT:

Daniel Doyle, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone: 301-415-3748, email: Daniel.Doyle@nrc.gov.

SUPPLEMENTARY INFORMATION:

Executive Summary

A. Need for the Regulatory Action

On May 23, 2025, President Donald J. Trump signed Executive Order (E.O.) 14300, “Ordering the Reform of the Nuclear Regulatory Commission.” Section 5, “Reforming and Modernizing the NRC's Regulations,” requires the NRC to undertake a review and wholesale revision of its regulations in title 10 of the Code of Federal Regulations (10 CFR) and guidance documents as guided by the policies set forth in section 2 of the E.O. This rulemaking addresses section 5(f) of E.O. 14300, which directs the NRC to “[e]stablish stringent thresholds for circumstances in which the NRC may demand changes to reactor design once construction is underway”; section 5(h) of E.O. 14300, which directs the NRC to “[a]dopt revised and, where feasible, determinate and data-backed thresholds to ensure that reactor safety assessments focus on credible, realistic risks”; and section 5(i) of E.O. 14300, which directs the NRC to “[r]econsider the regulations governing the time period for which a renewed license remains effective, and extend that period as appropriate based on available technological and safety data.” Additionally, as part of the NRC's overarching review of all of its regulations, the agency identified a number of additional changes to the NRC's regulations that will improve the efficiency and efficacy of its licensing process that are also included in this rulemaking.

These changes are the culmination of decades of combined experience, feedback from nuclear experts, lessons learned by the NRC and industry, international experience, and prior efforts to modernize the regulatory framework. They have undergone thoughtful preparation and internal vetting by the NRC technical experts. While some changes in this rule had not been advanced as regulatory priorities, the direction in the E.O. catalyzed agency efforts to accelerate modernization of 10 CFR part 50, “Domestic Licensing of Production and Utilization Facilities,” and 10 CFR part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” and make conforming changes to 10 CFR part 53, “Risk-Informed, Technology-Inclusive Regulatory Framework for Commercial Nuclear Plants.” Therefore, this proposed rule aims to modernize reactor licensing, safety oversight, and siting practices addressing sections 5(f), 5(h), and 5(i) of E.O. 14300 and additional items that would contribute to adding generation to the electrical grid. In developing the proposed changes, the NRC has considered the benefits of increased availability of, and innovation in, nuclear power to our economic and national security consistent with section 501(a) of the Accelerating Deployment of Versatile, Advanced Nuclear for Clean Energy Act of 2024 (ADVANCE Act) and section 3 of E.O. 14300.

B. Major Provisions

Major provisions of this proposed rule include changes in the following areas:

Expedited Construction of Certain Structures, Systems, and Components

The NRC is proposing to amend its regulations applicable to the definition of construction in 10 CFR 50.10, “License required; limited work authorization,” 10 CFR 51.4, “Definitions,” and 10 CFR 53.020, “Definitions,” in order to focus the scope of activities that are considered construction on structures, systems, and components (SSCs) for which construction can affect attributes of the SSC material to the SSC's capability to perform a safety-significant function and thereby reduce the cost impact of the current definition of construction. Additionally, the NRC proposes to include new paragraphs in 10 CFR 50.10(h) and 53.1130(e), which would issue a general license for beginning construction upon docketing an application for a license that would authorize construction of a nuclear plant, subject to conditions that ensure safety, security, and appropriate environmental review.

Determinate and Data-Backed Thresholds for Reactor Safety Assessments

The NRC is proposing to amend its regulations by revising 10 CFR 50.2, “Definitions,” to add the terms “design basis event” (DBE) and “beyond design basis event” (BDBE). The proposed changes also include a conforming revision to the definition of DBE in 10 CFR 50.49, “Environmental

Removal of IEEE-323-1974 Reference in Footnote 3 of 10 CFR 50.49

The NRC is proposing to revise the regulations in 10 CFR 50.49 to delete footnote 3, which clarifies that safety-related electric equipment is referred to as Class 1E equipment in Institute of Electrical and Electronics Engineers (IEEE) Standard 323-1974. The reference in this footnote is now unnecessary because the NRC has established the connection between “safety-related” electric equipment and “Class 1E” equipment elsewhere.

Expanded Alternative Requests Under 10 CFR 50.55a(z)

The NRC is proposing to amend its regulations to allow licensees to request a broader scope of alternatives to the requirements in 10 CFR 50.55a, “Codes and standards.” Currently, alternatives under 10 CFR 50.55a(z), “Alternatives to codes and standards requirements,” are limited to the requirements in 10 CFR 50.55a(b), “Use and conditions on the use of standards,” through (h), “Protection and safety systems.” Expanding the scope of alternatives permitted under 10 CFR 50.55a to all requirements in 10 CFR 50.55a would allow for added flexibility without requiring exemptions, while relying on the well-understood existing criteria of acceptable level of quality and safety (10 CFR 50.55a(z)(1)) and hardship without a compensating increase in quality or safety (10 CFR 50.55a(z)(2)) for consistent and predictable regulatory outcomes. This proposed action would allow nuclear power plant licensees and applicants for construction permits (CP), operating licenses (OL), combined licenses (COL), standard design certifications, standard design approvals, and manufacturing licenses (ML) to request authorization of voluntary alternatives to a broader scope of requirements in 10 CFR 50.55a.

Risk-Informing 10 CFR 50.59 and Allowing Flexibility for Changes to Methods

There are two proposed changes to 10 CFR 50.59, “Changes, tests, and experiments.” The first proposed change to the regulation would allow the use of quantitative risk results to demonstrate a change to the facility does not result in a “more than minimal increase” as the phrase is used in 10 CFR 50.59(c)(2)(i) and (ii). The second proposed change would allow licensees to make changes to methods that would previously have required NRC review under 10 CFR 50.59(c)(2)(viii) or 53.1550(a)(2)(iv), provided the licensee adopts an acceptable verification, validation, and uncertainty quantification (VVUQ) program in accordance with a proposed new 10 CFR 50.221, “Credibility requirements for modeling and simulation.” This shift would enable licensee-led evaluations and allow for the evaluation of advanced modeling methods through structured processes rather than fixed requirements.

Minimum Decommissioning Funding Assurance for Non-Large Light-Water Reactors

The NRC is proposing rule changes in 10 CFR 50.75, “Reporting and recordkeeping for decommissioning planning,” to allow certain new reactor applicants and licensees to submit a design-specific decommissioning cost estimate that is less than the approved table of minimum amounts ( i.e., “formula”) values in 10 CFR 50.75(c). The current regulations restrict minimum funding assurance for decommissioning to the table of minimum amounts values or greater. The NRC is proposing similar changes to 10 CFR part 53, which currently only allows for a site-specific decommissioning cost estimate as the certification amount. These changes would more broadly accommodate new reactor technologies.

Incorporation of Streamlined Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants

This proposed rule would incorporate an appendix T, “Streamlined Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,” to 10 CFR part 50. The proposed appendix T to 10 CFR part 50 would provide a voluntary alternative to appendix B, “Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,” to 10 CFR part 50 that applicants, who meet certain conditions, could use in their respective applications. This action would support implementation of E.O. 14300, section 5, through (1) the establishment of performance-based quality assurance (QA) criteria that provide explicit direction on the use of a graded approach for applying QA requirements to SSCs relative to their safety and risk contributions to the overall nuclear facility; (2) the incorporation of QA terminology and methodologies used across various industries; (3) enhanced regulatory certainty during the application process; and (4) the removal, when appropriate, of NRC oversight of suppliers and vendors of products and services related to nuclear power plant and fuel reprocessing plant SSCs subject to this proposed rule.

Updates to Construction Permit Requirements and Related Licenses

The proposed rule would revise the content of applications relevant to technical information required for CPs in 10 CFR 50.34, “Contents of applications; technical information,” to remove some overly-prescriptive wording; clarify that the level of detail provided in a preliminary safety analysis report should be sufficient to permit the NRC to make the findings in 10 CFR 50.35(a), 10 CFR 50.40, “Common standards,” and 10 CFR 50.50, “Issuance of licenses and construction permits”; and adjust some of the light-water reactor (LWR)-centric language to be more technology-inclusive. Conforming changes to parallel sections in 10 CFR part 52 are also proposed.

Alternative Risk-Informed and Performance-Based Acceptance Criteria for 10 CFR Parts 50 and 52

The NRC is proposing to amend its regulations by adding new, standalone provisions in 10 CFR 50.220 and 10 CFR 52.220, entitled “Use of risk-informed and performance-based alternatives to acceptance criteria.” These provisions would allow licensees and applicants to voluntarily submit and use technology-inclusive, risk-informed or performance-based acceptance criteria as alternatives to existing prescriptive requirements.

In addition, the NRC is proposing revisions to appendix A to 10 CFR part 50 to update and clarify its regulations for General Design Criteria (GDC). The proposed changes would explicitly allow deviations from the GDCs without requiring exemptions and would revise GDC 28, “Reactivity limits,” to remove the prescriptive requirement to evaluate control rod ejection and drop accidents. Instead, applicants would be permitted to propose an alternative design basis accident for reactivity control systems.

Establishing Thresholds for Changes to Reactor Designs During Construction and Operation Under 10 CFR Parts 52 and 53

In response to E.O. 14300, section 5(f), the NRC is proposing to amend its regulations related to reactor design changes made during construction and operation under the 10 CFR parts 52 and 53 licensing approaches. These changes would impact licensees that reference a certified design or manufacturing license under 10 CFR part 52 or 53. The objectives of these proposed changes are to establish appropriate thresholds for NRC-initiated changes as well as provide additional flexibility and reduce unnecessary regulatory burden for licensee-initiated changes.

Revision of the Emergency Preparedness Regulations for Nuclear Power Reactors

The NRC is proposing to amend its regulations to create adaptable licensing pathways for emergency preparedness (EP). Consistent with E.O. 14300, the NRC's objectives for this proposed rule are to streamline the licensing process, provide regulatory certainty for the deployment of new reactor technologies, and remove prescriptive language of lesser safety significance for licensed facilities. Central to these proposed changes is a strengthened, more risk-informed approach to EP for providing reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency.

Optional Submittal of Operational Programs

This rulemaking would allow a developer the option to voluntarily submit operational programs for NRC review and approval with an ML application. The intent is to allow construction permit/operating license (CP/OL) and COL applicants the flexibility to reference the standardized programs approved in the ML, use their own approved programs, or use a combination of both. Early review of these programs would support streamlined CP/OL or COL reviews.

Early Site Permit for Nuclear Power Plants

The NRC is proposing to amend its regulations by revising the provisions applicable to early site permit (ESP) licensing and approval processes for nuclear power plants. These amendments would eliminate the requirement for an ESP expiration date, clarify the applicability of various requirements to ESPs, and propose necessary conforming amendments throughout the NRC's regulations to enhance the NRC's necessary regulatory effectiveness and efficiency in implementing its licensing and approval processes.

Manufacturing License Term Extension

The NRC is proposing to amend the regulations in 10 CFR 52.173, “Duration of manufacturing license,” and 52.181, “Duration of renewal,” to change the duration of an ML to 40 years and the duration of the renewed ML to 40 years. By amending the regulations with these proposed changes, the ML would be consistent with the durations for certified designs, thereby increasing efficiency in building new reactors.

Nuclear Power Plant License Renewal

The NRC is proposing to amend its regulations for renewing nuclear power plant OLs. The revisions would extend the duration of renewed licenses, allow applicants to voluntarily propose alternative risk-informed and performance-based criteria, and remove several prescriptive requirements related to the application process and post-approval recordkeeping. These changes would enhance regulatory flexibility and efficiency to facilitate operational extensions for the current nuclear fleet.

Enhancing Flexibility of Reactor Site Criteria

The existing regulatory framework requires all stationary power reactor applications submitted after January 10, 1997, to follow the siting criteria in subpart B to 10 CFR part 100, “Evaluation Factors for Stationary Power Reactor Site Applications on or After January 10, 1997,” without consideration of reactor type, size, output, radiological consequence, or other factors that can widely vary given the breadth of power reactor designs considered for future construction and deployment in the United States. Subpart A, “Evaluation Factors for Stationary Power Reactor Site Applications Before January 10, 1997 and for Testing Reactors,” to 10 CFR part 100, “Reactor Site Criteria,” provides less prescriptive regulatory requirements for reactor siting but only applies to power reactor applications submitted prior to January 10, 1997, or an application for a testing reactor, as defined in 10 CFR 50.2. To increase the flexibility of regulatory requirements for the full spectrum of prospective reactor technologies, including non-stationary reactors, this proposed change would (1) revise subpart A to 10 CFR part 100 to include Tier 1 power reactors, as defined in proposed 10 CFR 100.3, “Definitions,” and DG-4036, “Graded Approach to Site Characterization for New Reactor Applications,” in addition to testing reactors; (2) remove appendix A to 10 CFR part 100 that applies to subpart A to 10 CFR part 100; and (3) revise subpart B to 10 CFR part 100 to include Tier 2 power reactors, which are those reactors that do not meet the entry criteria for subpart A to 10 CFR part 100. This proposed change will be accompanied by draft guidance on application content, including the entry criteria for subpart A to 10 CFR part 100, a clarification on site parameters to be included in a site parameter envelope, and an explanation of a graded approach to site characterization for all external hazards to be considered under both subparts A and B to 10 CFR part 100. In addition, the NRC proposes to revise 10 CFR part 100 to maintain the agency's long-standing preference for siting reactors in areas of low population density, while providing flexibility to allow siting reactors in areas of greater population density when justified by an assessment comparing the societal risks and societal benefits of siting reactors in those areas. Implementing guidance for these assessments will be developed.

Increased Enrichment of Conventional and Accident Tolerant Fuel Designs for Light-Water Reactors

The NRC is proposing to amend its regulations related to the use of conventional and accident tolerant LWR fuel designs. The NRC's goal is to establish effective and efficient licensing of the use of fuels enriched to greater than 5.0 weight percent uranium-235 while continuing to provide reasonable assurance of adequate protection of public health and safety. The new requirements also would address fuel fragmentation, relocation, and dispersal in relation to the key accident tolerant fuel components of increased enrichment and burnup limits.

C. Cost and Benefits

The NRC prepared a draft regulatory analysis to determine the expected quantitative and qualitative costs of the proposed rule and associated guidance. The draft regulatory analysis concluded that the proposed rule and associated guidance would result in undiscounted total net savings of $1.86 billion to the NRC and industry ($802.10 million using a 7 percent discount rate and $1.26 billion using a 3 percent discount rate).

The draft regulatory analysis also considers qualitative factors to be considered in the NRC's rulemaking decision. Qualitative factors include

For more information, the draft regulatory analysis is available as indicated in the “Availability of Documents” section of this document.

Table of Contents

I. Obtaining Information and Submitting Comments A. Obtaining Information B. Submitting Comments II. Executive Order 14300: Ordering the Reform of the Nuclear Regulatory Commission III. Background—Expedited Construction of Certain Structures, Systems, and Components A. Definition of Construction B. General Licenses IV. Discussion—Expedited Construction of Certain Structures, Systems, and Components A. Definition of Construction B. Safety Review C. Environmental Review D. General Licenses and Generic Finality V. Background—Determinate and Data-Backed Thresholds for Reactor Safety Assessments VI. Discussion—Determinate and Data-Backed Thresholds for Reactor Safety Assessments VII. Background—Removal of IEEE-323-1974 Reference in Footnote 3 of 10 CFR 50.49 VIII. Discussion—Removal of IEEE-323-1974 Reference in Footnote 3 of 10 CFR 50.49 IX. Background—Expanded Alternative Requests Under 10 CFR 50.55a(z) X. Discussion—Expanded Alternative Requests Under 10 CFR 50.55a(z) XI. Background—Risk-Informing 10 CFR 50.59 and Allowing Flexibility for Changes to Methods A. Use of Quantitative Risk Results B. Improved Flexibility for Changes to Methods of Evaluation XII. Discussion—Risk-Informing 10 CFR 50.59 and Allowing Flexibility for Changes to Methods A. Use of Quantitative Risk Results B. Improved Flexibility for Changes to Methods of Evaluation XIII. Background—Minimum Decommissioning Funding Assurance for Non-Large Light-Water Reactors XIV. Discussion—Minimum Decommissioning Funding Assurance for Non-Large Light-Water Reactors XV. Background—Incorporation of Streamlined Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants A. Historic Quality Assurance Requirements Perspectives and Emergent Issues B. NRC Responses to These Issues XVI. Discussion—Incorporation of Streamlined Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants A. Introduction and Scope B. Definitions C. General Requirements D. Quality Assurance Requirements E. Quality Assurance for Software Used in Design and Analysis, and Digital Items Important to Safety F. Proposed Conforming Changes to 10 CFR 50.4, 50.34, 50.54, 50.55, 52.79, 53.020, 53.040, 53.460, 53.500, 53.865, 53.1309, 53.1369, 53.1416, and 53.1565 XVII. Background—Updates to Construction Permit Requirements and Related Licenses XVIII. Discussion—Updates to Construction Permit Requirements and Related Licenses XIX. Background—Alternative Risk-Informed and Performance-Based Acceptance Criteria for 10 CFR Parts 50 and 52 A. Need for Regulatory Flexibility B. Enabling Risk-Informed and Performance-Based Alternatives XX. Discussion—Alternative Risk-Informed and Performance-Based Acceptance Criteria for 10 CFR Parts 50 and 52 XXI. Background—Establishing Thresholds for Changes to Reactor Designs During Construction and Operation Under 10CFR Parts 52 and 53 A. Development of Tiers of Information and Processes for Changes and Departures in Design Certification Rules B. Licensing Experience and Improvement Initiatives Regarding Information Designation and Change Processes for Design Certifications C. Severe Accidents XXII. Discussion—Establishing Thresholds for Changes to Reactor Designs During Construction and Operation Under 10 CFR Parts 52 and 53 A. Standardization B. Definitions C. Processes for Changes and Departures XXIII. Background—Revision of the Emergency Preparedness Regulations for Nuclear Power Reactors A. Existing Emergency Preparedness Frameworks for Nuclear Power Reactors B. Protective Actions and Emergency Planning Zones XXIV. Discussion—Revision of the Emergency Preparedness Regulations for Nuclear Power Reactors A. Emergency Plan Licensing Flexibility B. Emergency Planning Zone Certainty C. Eliminating Redundant Requirements D. Risk-Informing the Emergency Plan Change Process XXV. Background—Optional Submittal of Operational Programs XXVI. Discussion—Optional Submittal of Operational Programs XXVII. Background—Early Site Permit for Nuclear Power Plants XXVIII. Discussion—Early Site Permit for Nuclear Power Plants XXIX. Background—Manufacturing License Term Extension XXX. Discussion—Manufacturing License Term Extension XXXI. Background—Nuclear Power Plant License Renewal XXXII. Discussion—Nuclear Power Plant License Renewal A. Extension of the Renewal Time Period B. Alternative Risk-Informed and Performance-Based Criteria C. Eliminate the Limitation on Early Application Submittal D. Eliminate the Required Application Content on Exemptions E. Eliminate the Required Application Content on Technical Specifications F. Reduce Ongoing, Post-Renewal Updates to the Final Safety Analysis Report XXXIII. Background—Enhancing Flexibility of Reactor Site Criteria XXXIV. Discussion—Enhancing Flexibility of Reactor Site Criteria XXXV. Background—Increased Enrichment of Conventional and Accident Tolerant Fuel Designs for Light-Water Reactors A. Accident Tolerant Fuels B. Rulemaking Development C. Background and History of Affected Regulations D. Regulatory Basis XXXVI. Discussion—Increased Enrichment of Conventional and Accident Tolerant Fuel Designs for Light-Water Reactors A. Criticality Accident Requirements in 10 CFR 50.68 B. Uranium Fuel Cycle Environmental Data—Table S-3 in 10 CFR 51.51 C. Environmental Effects of Transportation of Fuel and Waste—Table S-4 in 10 CFR 51.52 D. Fissile Material Packaging Requirements in 10 CFR 71.55 E. Control Room Requirements in 10 CFR 50.67 and GDC 19 F. Fuel Dispersal XXXVII. Specific Questions XXXVIII. Regulatory Flexibility Certification XXXIX. Regulatory Analysis XL. Backfitting and Issue Finality XLI. Cumulative Effects of Regulation XLII. Plain Writing XLIII. National Environmental Policy Act XLIV. Paperwork Reduction Act XLV. Executive Orders A. Executive Order 12866: Regulatory Planning and Review (as Amended by Executive Order 14215, Ensuring Accountability for All Agencies) B. Executive Order 14154: Unleashing American Energy C. Executive Order 14192: Unleashing Prosperity Through Deregulation D. Executive Order 14267: Reducing Anti-Competitive Regulatory Barriers XLVI. Voluntary Consensus Standards XLVII. Availability of Guidance XLVIII. Availability of Documents I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID NRC-2025-0975 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-2025-0975.

• 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 ADAMS Public Search.” For problems with ADAMS, please contact the NRC's Public Document Room (PDR) reference staff at 1-800-397-4209, at 301-415-4737, or by email to 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 of this document.

• 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.

• Technical Library: The Technical Library, which is located at Two White Flint North, 11545 Rockville Pike, Rockville, Maryland 20852, is open by appointment only. Interested parties may make appointments to examine documents by contacting the NRC Technical Library by email at Library.Resource@nrc.gov between 8 a.m. and 4 p.m. Eastern Time, Monday through Friday, except Federal holidays.

• Public Meeting: The NRC will conduct public meetings to describe the proposed amendments and answer questions from the public on the proposed rule. The NRC will publish a notice of the location, time, and agenda of the meetings on the NRC's public meeting website within 10 calendar days of the meetings. Stakeholders should monitor the NRC's public meeting website for information about the public meetings at: https://www.nrc.gov/public-involve/public-meetings/index.cfm.

B. Submitting Comments

Comments must be submitted using https://www.regulations.gov by 11:59 p.m. Eastern Time on August 31, 2026. Please include Docket ID NRC-2025-0975 in your comment submission.

The NRC cautions you not to include identifying or contact information that you do not want to be publicly disclosed in your comment submission. The NRC will post all comment submissions at 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. Executive Order 14300: Ordering the Reform of the Nuclear Regulatory Commission

On May 23, 2025, President Donald J. Trump signed Executive Order (E.O.) 14300, “Ordering the Reform of the Nuclear Regulatory Commission.” Section 5, “Reforming and Modernizing the NRC's Regulations,” requires the Nuclear Regulatory Commission (NRC) to undertake a review and wholesale revision of its regulations and guidance documents as guided by the policies set forth in section 2 of the E.O. This rulemaking addresses section 5(f), which directs the NRC to “[e]stablish stringent thresholds for circumstances in which the NRC may demand changes to reactor design once construction is underway”; section 5(h), which directs the NRC to “[a]dopt revised and, where feasible, determinate and data-backed thresholds to ensure that reactor safety assessments focus on credible, realistic risks”; and section 5(i), which directs the NRC to “[r]econsider the regulations governing the time period for which a renewed license remains effective, and extend that period as appropriate based on available technological and safety data.” Additionally, as part of the NRC's overarching review of all of its regulations, the agency identified a number of additional changes to the NRC's regulations that will improve the efficiency and efficacy of its licensing process that are also included in this rulemaking. In developing the proposed changes, the NRC has considered the benefits of increased availability of, and innovation in, nuclear power to our economic and national security consistent with section 501(a) of the Accelerating Deployment of Versatile, Advanced Nuclear for Clean Energy Act of 2024 (ADVANCE Act) and section 3 of E.O. 14300.

III. Background—Expedited Construction of Certain Structures, Systems, and Components

A. Definition of Construction

Section 185 of the Atomic Energy Act of 1954, as amended (AEA), requires that the NRC grant construction permits (CPs) to applicants for licenses to construct or modify production or utilization facilities, if the applications for such permits are acceptable to the NRC. However, the term “construction” is not defined anywhere in the AEA. Instead, construction is defined within section 50.10, “License required; limited work authorization” of part 50, “Domestic Licensing of Production and Utilization Facilities,” in title 10 of the Code of Federal Regulations (10 CFR). The Commission last updated this definition in 2007 as part of the limited work authorization (LWA) final rule, “Limited Work Authorizations for Nuclear Power Plants” (72 FR 57416; October 9, 2007).

In developing this definition of construction in the 2007 LWA final rule, the Commission concluded that the definition of construction should parallel the agency's jurisdiction because activities outside the definition of construction would not require prior authorization. Therefore, the Commission “determined that construction should include all of the activities that have a reasonable nexus to radiological health and safety, or common defense and security” (72 FR 57429; October 9, 2007). For the 2007 LWA final rule, the scope of structures, systems, and components (SSCs) falling within the definition of construction was derived from the scope of SSCs that are included in the program for monitoring the effectiveness of maintenance at nuclear power plants, as defined in 10 CFR 50.65(b), because “the definition is well understood and there is good agreement on its implementation” (72 FR 57429-30; October 9, 2007). The SSCs were supplemented with those necessary to comply with emergency preparedness and security regulations because they also have a reasonable nexus to radiological safety or are required for the common defense and security.

New reactor designs and deployment strategies, as well as lessons learned from previous examples, warrant a fresh look at how application of the 10 CFR 50.10 definition of construction may unnecessarily restrict construction of nuclear power plants using modern construction techniques on optimized schedules. Additionally, some prospective applicants to construct a nuclear power reactor have indicated

B. General Licenses

In regard to general licenses for utilization facilities, section 109 of the AEA, “Component and Other Parts of Facilities,” authorizes the Commission to “issue general licenses for domestic activities required to be licensed under section 101 [of the AEA]” with respect to those utilization facilities determined by the Commission under section 11cc.(2) of the AEA. Section 11cc.(2) of the AEA defines such facilities as “any important component part especially designed for [a utilization facility as defined under sec. 11cc.(1) of the AEA] as determined by the Commission.” Under section 109 of the AEA, the Commission may issue a general license authorizing construction of such “important component parts” if it “determines in writing that such general licensing will not constitute an unreasonable risk to the common defense and security.”

In order to accommodate the business models for new reactor designs some prospective vendors are proposing and to enable rapid deployment strategies for advanced reactor technologies, the NRC plans to revise its regulations to allow the use of general licenses for construction of important component parts of a utilization facility.

IV. Discussion—Expedited Construction of Certain Structures, Systems, and Components

A. Definition of Construction

The NRC is proposing to revise its regulations to update the definition of “construction” in 10 CFR 50.10, 10 CFR 51.4, “Definitions,” and 10 CFR 53.020, “Definitions,” to facilitate the safe construction of nuclear power plants using modern construction techniques on optimized schedules.

Specifically, this proposed rule would revise the definition of construction to include SSCs for which construction can affect the SSC's capability to perform a safety-related or safety-significant function and will, therefore, require NRC approval before commencing construction. An additional purpose of this change is to afford license applicants, when justified, additional flexibility to build or install SSCs whose safety-related or safety-significant functions are not significantly affected by those activities at a site prior to the issuance of a license. Those SSCs that are constructed without NRC authorization may still be subject to additional operational requirements as part of any subsequent operating license (OL) that would be issued.

B. Safety Review

Since the issuance of the 2007 LWA final rule, the NRC has observed that prospective advanced reactor applicants have designed their facilities with separation between nuclear and balance of plant SSCs in mind, such that many of the criteria in the current definition for a construction activity are not met for certain SSCs. Some stakeholders have maintained that such SSCs do not have a reasonable nexus to safety and therefore the unmet criteria are not necessary to provide reasonable assurance of adequate protection to the health and safety of the public; therefore, those unmet criteria are not needed or do not serve the underlying purpose of the rule. Lacking a revision to the definition, the remaining unmet criteria prevent prospective applicants from constructing such components without first obtaining an LWA, CP, a COL, or an exemption.

The proposed construction definition would be limited to those SSCs for which construction activities may have a significant impact on radiological health and safety. For other SSCs, even those that may have a nexus to radiological health and safety during operation, operational requirements should suffice, and the NRC need not license the construction of those SSCs.

The flexibility afforded by this proposed change would rely on an applicant-performed analysis and categorization of the SSCs of the facility to those that do and those that do not meet the definition of construction.

The set of SSCs that would meet the definition of construction should include only those SSCs that perform safety-related functions or that perform safety-significant functions and the successful completion of those functions may be impacted by construction. For those SSCs, prior NRC approval for construction would be required because inadequate design or construction of those SSCs could have a substantial contribution to radiological risk during operation.

Activities undertaken to build onsite emergency facilities necessary to comply with either 10 CFR 50.160, “Emergency preparedness for small modular reactors, non-light-water reactors, and non-power production or utilization facilities,” or 10 CFR 50.47, “Emergency plans,” and appendix E to 10 CFR part 50, “Emergency Planning and Preparedness for Production and Utilization Facilities,” or 10 CFR 53.855, “Emergency preparedness,” as applicable, would not be considered SSCs that meet the definition of construction. Historically, emergency response facilities (ERF) were included due to their reasonable nexus to radiological health and safety, but they would not fall under the criteria in proposed 10 CFR 50.10(a)(1)(i) through (iii) or 10 CFR 53.020. Instead, applicants would comply with the requirements of 10 CFR 50.160, or 10 CFR 50.47 and appendix E to 10 CFR part 50, or 10 CFR 53.855, as applicable. The applicant would need to be aware of all functional requirements of 10 CFR 50.160, or 10 CFR 50.47 and appendix E to 10 CFR part 50, or 10 CFR 53.855 for ERFs. These functional requirements would need to be validated in a preoperational exercise, which would satisfy the historical reasons why ERFs were previously included in the construction definition.

For license applicants under 10 CFR parts 50 and 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” the proposed revisions to the definition of construction would also support use of the SSC categorization methodology for designs licensed under the technology-inclusive, risk-informed, and performance-based methodology described in regulatory guide (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,” dated June 2020. Under this framework, the applicant would use its probabilistic risk assessment (PRA) of the design to analyze the function of the SSCs. As explained in RG 1.233, the applicant's analysis would result in the classification of SSCs into one of four categories: “safety-related,” “non-safety-

Applicants under 10 CFR part 50 or 52, when determining whether SSCs fall under the criteria in proposed 10 CFR 50.10(a)(1)(i) through (iii), would confirm that other considerations do not require that the resulting list of SSCs should otherwise be subject to NRC quality assurance (QA) requirements for design in appendix B to 10 CFR part 50, “Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants.” Criterion III of appendix B to 10 CFR part 50 sets forth requirements for design control, and appendix B to 10 CFR part 50 includes other requirements ( e.g., for records and audits) that apply to the design of SSCs subject to appendix B to 10 CFR part 50.

In addition, applicants under 10 CFR parts 50 and 52 would also confirm that other considerations would not require that any other SSC should be subject to general design criterion (GDC) 1, “Quality standards and records,” in appendix A to 10 CFR part 50. Criterion 1 of appendix A to 10 CFR part 50 also imposes corresponding requirements for SSCs important to safety but not safety-related to the extent such requirements are commensurate with an SSC's importance to safety. Safety-related SSCs are subject to all requirements in appendix B to 10 CFR part 50, including QA requirements applicable to facility operation.

Application of the QA requirements for design in proposed appendix T, “Streamlined Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,” to 10 CFR part 50 (see section XVI of this document), could also affect the definition of construction for applicants under 10 CFR part 50, 52, or 53, “Risk-Informed, Technology-Inclusive Regulatory Framework for Commercial Nuclear Plants.” Specifically, the definition of construction proposed in this rule would include SSCs subject to the requirements of proposed appendix T to 10 CFR part 50, and, in certain instances, the structures built for those SSCs.

Applicants under 10 CFR parts 50 and 52 must comply with criterion 1 of appendix A to 10 CFR part 50. This criterion requires that “SSCs important to safety be designed, fabricated, erected, and tested” to QA standards commensurate with the importance of the safety functions to be performed. However, an SSC could be subject to performance requirements for operation but would not warrant the application of QA measures under GDC 1 such as QA for design. Specifically, the SSC could be commercial grade but also subject to specified operational performance requirements. Such an SSC could be constructed without a license. The rationale for the approach rests on the fact that if no NRC QA requirement for design, fabrication, erection, and testing applies to an SSC, then there is nothing uniquely related to nuclear safety for the NRC to approve with respect to construction of the SSC. Further, design limits on the operation of the SSCs that do perform safety functions would prevent or mitigate the safety effects of the failure of SSCs not subject to NRC QA requirements applicable to construction. Accordingly, construction of such SSCs would not have a reasonable nexus to nuclear safety, even if operation of the SSC did have a reasonable nexus to nuclear safety.

The failure of an SSC that has an effect on the safety of operation could warrant operational requirements with respect to SSCs that perform safety functions in response to such failures and possibly operational requirements with respect to the SSCs themselves.

An applicant's designation of an SSC as something that does not meet the definition of construction would not restrict the NRC from imposing such operational requirements to address radiological health and safety.

The NRC encourages pre-application engagement when licensees plan to undertake significant preconstruction activities to assist the NRC staff to further understand which SSCs have a reasonable nexus to radiological health and safety at the construction phase. An applicant may communicate its plan to comply with the regulations, including an SSC classification methodology to the NRC as part of preapplication interactions. The NRC staff will provide feedback as appropriate. Alternatively, one or more prospective license applicants could propose a generic SSC classification methodology, which the NRC staff could review and endorse as acceptable guidance outside a particular licensing action. The approved methodology could then be referenced by multiple applicants.

Similar to the existing regulations, if an applicant under the proposed rule determines that an SSC falls within the scope of the definition of construction in the proposed 10 CFR 50.10 or 53.020, an exemption request or an LWA would need to be submitted to the NRC to allow for this construction activity to occur prior to issuance of a CP or COL. Similar to the existing processes, the LWA would be granted if the underlying requirements of 10 CFR 50.10(d) and (e) or 10 CFR 53.1130, “Limited work authorizations, general licenses,” as applicable, are met.

C. Environmental Review

As explained in section IV.A., “Definition of Construction,” of this document, a CP, COL, or LWA applicant does not need to obtain an NRC license to build SSCs that do not meet the definition of construction in 10 CFR 51.4. As long as there is no other Federal action authorizing these activities, under 10 CFR 51.20 through 51.22, an environmental review under the National Environmental Policy Act (NEPA) is not required. Activities undertaken to build SSCs excluded from the definition of construction are not part of an NRC licensing action because such construction activities do not have a reasonable nexus to nuclear safety, even if operation of a particular SSC does have a reasonable nexus to nuclear safety. An applicant's classification of an SSC as not safety-significant, that is built prior to issuance of a license, would not restrict the NRC from imposing operational requirements on those SSCs through a later action.

For SSCs that do meet the definition of construction where the NRC would authorize construction, the NRC must perform an environmental review in accordance with 10 CFR 50.10 or 53.610, “Construction,” as applicable,

D. General Licenses and Generic Finality

The NRC proposes to add new 10 CFR 50.10(h) and 53.1130(e), which would implement the authority in section 109 of the AEA to create a general license allowing construction of an important component part of a specified class of commercial nuclear plants. The specified class of plants would be those plants of a design the NRC previously approved in a licensing action in which the NRC also granted “generic finality,” as discussed later, and for which operation has been authorized. The important component part defined as a utilization facility under Section 11cc.(2) of the AEA and subject to the general license would be the portion of the plant constructed on site except for the reactor vessel, the reactor coolant system, and associated reactivity control and heat removal systems. The general license would authorize construction of the important component part upon docketing of an application for a license that would authorize construction of the nuclear plant, subject to conditions. The conditions would provide reasonable assurance of adequate protection of the health and safety of the public and common defense and security, and would also ensure an appropriate level of environmental review. The NRC also proposes a conforming change to 10 CFR 50.10(c) and 53.610(b) to provide that construction may occur under the general licenses issued in 10 CFR 50.10(h) and 53.1130(e), respectively.

The construction activities authorized by the general license would be limited to those SSCs for which a previously approved design was provided generic finality, but would not include the reactor vessel, the reactor coolant system, and associated reactivity control and heat removal systems. The NRC expects that the construction of SSCs approved through a general license would be for those SSCs that are not inherently sensitive to the site-specific characteristics of a proposed deployment site. Although construction of many SSCs in a particular design could conceptually be approved through a general license, the NRC cannot approve the construction of an entire utilization facility through this provision.

The proposed new regulations would enable future applicants to reference previously reviewed and approved information only when significant safety and environmental issues related to design, construction, and operation are generically resolved in a manner that applies to the intended use of the information. For example, to qualify for a general license, an applicant would have to reference a nuclear reactor design that was afforded generic finality by the NRC and successfully constructed under NRC oversight and placed into operation. Also, the applicant's proposed site would have to fall within the corresponding site parameter envelope that was provided in the request for generic finality. Therefore, an application that satisfies the proposed new regulation would provide reasonable assurance of adequate protection of public health and safety and common defense and security equivalent to satisfaction of existing regulations, and there would have been a prior hearing opportunity on the reactor design being referenced. Further, the proposed regulation would require the general licensee to allow for NRC inspections that the Commission deems necessary related to activities performed under the general license.

The general license regulation would also include conditions to address environmental considerations. The OL or COL (as applicable) of the plant for which generic finality was approved would either have met the criteria for categorical exclusion or had a finding of no significant impact after preparation of an environmental assessment. Provided that the environmental characteristics of the proposed plant fall within the environmental parameters for the plant for which generic finality was approved, the environmental effects of a subsequent plant would not exceed those of the approved plant and would be acceptable. In addition, the applicant proposing to use the general license would have to propose a plan for redress of any adverse environmental impact from conduct of activities under the general license should such redress be necessary. This proposed requirement would be similar to the requirements in 10 CFR 50.10(d)(3)(iii), which requires a redress plan as part of an application for an LWA, and 10 CFR 50.12(b)(2), which requires the Commission to consider redress of adverse environmental impacts in determining whether to grant an exemption permitting the conduct of construction activities prior to the issuance of a CP.

The proposed general license regulation would also require that the general licensee has notified the NRC that all applicable permits, licenses, approvals, and other entitlements in connection with the proposed action that the general licensee was responsible for obtaining have been obtained. In addition, the proposed general license would require that applicable Federal environmental consultations have been completed. This would ensure that construction activities would not begin unless the NRC has the information it would need to fulfill its obligations for environmental review under the AEA, NEPA, and other relevant laws.

In addition, the proposed general license regulation would clarify that any activities undertaken by the general licensee or on its behalf under the general license would be entirely at the risk of the general licensee and would have no bearing on the issuance of a license with respect to the requirements of the AEA, and rules, regulations, or orders issued under the AEA. However, the general licensee would be able to mitigate this additional regulatory risk through careful site selection to ensure that site characteristics are within the bounds of the postulated site parameters and by performing construction activities following appropriate QA and fitness-for-duty programs.

Based on the proposed general license requirements in 10 CFR 50.10(h) and 53.1130(e), the Commission has determined that such general licensing would be for only parts of utilization facilities, not constitute an unreasonable risk to the common defense and security, and, therefore, be consistent with the authority provided to the Commission by section 109a. of the AEA.

In addition, in order to facilitate the use of the general licensing concept, the NRC proposes to add conforming changes to the following regulations.

The NRC proposes to add new 10 CFR 50.34(b)(14) which would require an OL application for those 10 CFR part 50 applicants that request the NRC to make a finding on generic finality, to include applicable site parameters postulated for the design, including the design-basis external hazard levels for the relevant external hazards, and an analysis and evaluation of the design in terms of those site parameters. Similarly, the NRC proposes to add new paragraph (bb) to 10 CFR 53.1369, “Contents of applications for operating licenses; technical information,” for 10 CFR part 53, “Risk-Informed, Technology-Inclusive Regulatory Framework for Commercial Nuclear Plants,” OLs for the same purpose. For COL applications under 10 CFR part 52 or 53, this application content requirement would be included in a new 10 CFR 52.79(a)(48) and 53.1416(i), respectively. The site parameters may be the same as or more severe than the site

The NRC also proposes to add a new 10 CFR 50.58(b)(7), which would require the Commission to include the request for generic finality as a proposed action in the notice of proposed action required by 10 CFR 2.105 for OL applications. Similarly, the NRC also proposes to add a new paragraph (b)(2) to 10 CFR 53.1375, “Review of applications,” which would require the Commission to include the request for generic finality as a proposed action in the notice of proposed action for a 10 CFR part 53 OL, required by 10 CFR 2.105. For COL applications, the NRC proposes to add a new 10 CFR 52.85(b) and 53.1422(b)(2), which would require the Commission to include the request for generic finality as a proposed action in the notice of hearing required by 10 CFR 2.104 for COL applications under 10 CFR parts 52 and 53, respectively. These changes would provide a hearing opportunity to the public on the request for generic finality. In addition, the Commission's ruling on a request for hearing or petition for leave to intervene under 10 CFR 2.309(d)(2) would consider that a petitioner may have an interest in the application if matters resolved in the licensing proceeding were to be afforded generic finality. This would enable petitioners whose property, financial, or other interests would not be directly affected by the issuance of the OL or COL for a particular reactor to have an opportunity to intervene on generic aspects of the design that would be afforded finality and would therefore not be subject to hearing if referenced in a later application that would affect the petitioner's property, financial, or other interest.

Consistent with the previous discussion, the NRC proposes to add new 10 CFR 50.57(d), which would permit the Commission to afford generic finality to generic aspects of the design of a utilization facility licensed under 10 CFR part 50, including postulated site parameters submitted pursuant to 10 CFR 50.34(b)(14), if it finds that the proposed generic design can be constructed and operated at sites having characteristics that fall within the site parameters postulated for the design. For the same reason and with the same conditions, the NRC proposes to add new paragraph (e) to 10 CFR 53.1387, “Issuance of operating licenses,” which would permit the Commission to afford generic finality to generic aspects of the design of a commercial nuclear plant licensed under 10 CFR part 53, including postulated site parameters submitted pursuant to 10 CFR 53.1369(bb).

Similarly, the NRC proposes to add new 10 CFR 52.97(d) which would permit the Commission to afford generic finality to generic aspects of the design of a utilization facility licensed under 10 CFR part 52, including postulated site parameters submitted pursuant to 10 CFR 52.79(a)(48), if it finds that the proposed generic design can be constructed and operated at sites having characteristics that fall within the site parameters postulated for the design. For the same reason and with the same conditions, the NRC proposes to add new paragraph (d) to 10 CFR 53.1440, “Issuance of combined licenses,” which would permit the Commission to afford generic finality to generic aspects of the design of a commercial nuclear plant licensed under 10 CFR part 53, including postulated site parameters submitted pursuant to 10 CFR 53.1416(i).

The regulations in 10 CFR 50.59, “Changes, tests and experiments,” that establish requirements for making changes to portions of the facility as described in the final safety analysis report (FSAR) for an OL are applicable to generic aspects of the design of a utilization facility that have been afforded generic finality because that design information would be included in the FSAR for the OL. Similarly, the regulations in 10 CFR 50.59 are applicable to generic aspects of the design of a utilization facility that are described in a COL FSAR and have been afforded generic finality.

Similarly, the regulations in 10 CFR part 53 that establish requirements for making changes to portions of the facility as described in the FSAR are applicable to generic aspects of the design of a utilization facility that have been afforded generic finality.

The NRC proposes to add new 10 CFR 50.58(b)(8) and 52.98(h) to include requirements to address finality for portions of 10 CFR part 50 OLs and 10 CFR part 52 COLs with respect to NRC reviews and hearings. Proposed 10 CFR 50.58(b)(8) would require the Commission to treat as resolved any issues referenced in following proceedings or in enforcement hearings (other than ones under 10 CFR 2.202(e)(1)) that were afforded finality pursuant to 10 CFR 50.57(d). The proposed 10 CFR 50.58(b)(8) would ensure that issues resolved in an approved request for generic finality (including, if applicable, the adequacy of a reactor design) are not re-adjudicated in the license proceedings where such information is referenced in the license applications. The proposed 10 CFR 52.98(h) would include substantially the same provisions for COLs with generic finality.

To address generic finality in 10 CFR part 53, the NRC proposes to add similar provisions to new paragraph (b) to 10 CFR 53.1390, “Finality of operating licenses,” and new paragraph (g) to 10 CFR 53.1443, “Finality of combined licenses.”

Proposed 10 CFR 53.1390(b) would require the Commission, in the proceedings for issuance of a CP, an OL, or a COL or in any enforcement hearing (other than one initiated under 10 CFR 53.1390(a)), to treat as resolved those matters resolved in the proceedings on the application or renewal of the referenced OL, including, if applicable, the adequacy of a reactor design where the referenced OL was afforded finality pursuant to 10 CFR 53.1387(e).

Proposed 10 CFR 53.1443(g) would require the Commission, in the proceedings for issuance of a CP, an OL, or a COL or in any enforcement hearing (other than one initiated under 10 CFR 53.1443(a)), to treat as resolved those matters resolved in the proceedings on the application or renewal of the referenced COL, including, if applicable, the adequacy of a reactor design where the referenced COL was afforded finality pursuant to 10 CFR 53.1440(d).

As written, the proposed generic finality provisions would allow the NRC to take appropriate action under the applicable backfitting or issue finality provision if the NRC determines that the generic finality approval or associated technical information presents safety concerns that warrant NRC action. As stated above, the generic finality that would be afforded under the proposed rule provisions would not apply to certain enforcement hearings. For example, proposed 10 CFR 50.58(d)(8) provides that finality would apply, in part, to “any enforcement hearing other than one initiated by the Commission under § 2.202(e)(1) of this chapter” (emphasis added). Enforcement hearings under 10 CFR 2.202(e)(1) are those which involve a backfit to modify a 10 CFR part 50 license, and 10 CFR 50.109 must be followed for such orders and the associated proceedings. Thus, generic finality would not apply in such cases so that the NRC could take appropriate action if the backfitting requirements in 10 CFR 50.109 are satisfied. Similarly, the proposed generic finality provisions in 10 CFR 52.98(h), 53.1390(b), and 53.1443(g) would provide that generic finality applies except in enforcement hearings initiated under the issue finality

V. Background—Determinate and Data-Backed Thresholds for Reactor Safety Assessments

Section 5(h) of E.O. 14300 directs the NRC to “[a]dopt revised and, where feasible, determinate and data-backed thresholds to ensure that reactor safety assessments are focused on credible, realistic risks.”

The NRC evaluated “reactor safety assessments,” focusing on assessments conducted to (1) demonstrate the capability of safety-related SSCs during design basis events (DBEs) ( e.g., as described in Chapter 15, “Transient and Accident Analysis,” of NUREG-0800, “Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition”), and (2) verify the ability of SSCs to withstand certain design basis conditions, including natural phenomena and environmental conditions ( e.g., high winds, seismic events, and conditions during normal operation and accident scenarios). To effectively address the direction in E.O. 14300, the NRC concluded that the most appropriate approach would be to clarify the terminology in 10 CFR 50.2, “Definitions,” to ensure that safety assessments are focused on credible, realistic risks.

The selection of DBEs and associated design basis parameters is a critical prerequisite for determining the safety of a nuclear facility. DBEs serve to identify the subset of SSCs subject to more stringent QA requirements and to establish the performance capabilities those SSCs must demonstrate under normal operation, anticipated operational events, and accident conditions. For example, 10 CFR 50.46, “Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors,” requires analysis of postulated loss-of-coolant accidents to verify the adequacy of emergency core cooling system (ECCS) designs. In addition to the general requirement to analyze SSC performance during DBEs, other regulations specify additional, event-specific accident analyses, often referred to as beyond design basis events (BDBEs). For example, 10 CFR 50.63, “Loss of all alternating current power,” provides requirements related to plants' abilities to withstand for a specified duration and recover from a station blackout (SBO).

The technical information required in applications for CPs and OLs, including the content of preliminary and FSARs, is outlined in 10 CFR 50.34, “Contents of applications; technical information.” Among other requirements, an applicant is required to evaluate siting considerations and the design and performance of SSCs that are intended to prevent accidents and mitigate their consequences. The regulatory processes described in 10 CFR part 52 include similar technical information requirements for the content of applications ( e.g., as specified in 10 CFR 52.17, “Contents of applications; technical information,” 52.47, “Contents of applications; technical information,” 52.79, “Contents of applications; technical information in final safety analysis report,” 52.137, “Contents of applications; technical information,” and 52.157, “Contents of applications; technical information in final safety analysis report”).

The technical information associated with the performance of safety assessments is documented in the preliminary or FSAR for CPs or OLs and COLs, respectively. The FSAR describes the evaluation methods used to establish design bases and perform safety analyses, the design and performance requirements for SSCs, and the methods used to demonstrate that those SSCs can perform their intended safety functions. Accordingly, the FSAR serves as an essential component of the licensing basis for a nuclear facility. It is also used to determine the appropriate regulatory process for licensing basis changes, such as those governed by 10 CFR 50.59, “Changes, tests and experiments,” or 10 CFR 50.90, “Application for amendment of license, construction permit, or early site permit.”

Within the power reactor licensing framework, the term “safety-related” is used to identify SSCs that require special treatment, including QA controls, environmental qualification, and compliance with applicable industry codes and standards. The current definition of “safety-related SSCs,” provided in 10 CFR 50.2, uses the term “design basis events” to define the scope of safety assessments needed to identify safety-related SSCs. However, 10 CFR 50.2 does not include a corresponding definition of “design basis event” or provide criteria for selecting events to be considered in the design basis. While 10 CFR 50.49, “Environmental qualification of electric equipment important to safety for nuclear power plants,” includes a definition of DBEs, that definition does not explicitly apply to the definition of safety-related SSCs in 10 CFR 50.2 and does not reference the use of determinate, data-backed thresholds.

Traditionally, the spectrum of DBEs used to identify safety-related SSCs, as defined in 10 CFR 50.2, has been based on information contained in Chapter 15 of NUREG-0800. While this approach had been effective for licensing large light-water reactors (LWRs) with designs similar to the currently operating power reactor fleet, the DBEs described in NUREG-0800 can have limited applicability to evolutionary LWR designs and non-LWR designs. For example, the lack of a more technology-inclusive definition for the term DBE in the current definition has created challenges with respect to clarity and reliability on the subset of SSCs that warrant special treatment. Furthermore, it has created the potential to require safety assessments that may not be focused on credible, realistic risks. Therefore, clarifying what constitutes a DBE based on determinate, data-backed thresholds would enhance the efficiency and consistency of future power reactor licensing reviews. Consequently, the NRC proposes to provide a definition for DBE in 10 CFR part 50.

Nonetheless, the NRC has not identified a need to propose a corresponding revision to the term “design bases” in 10 CFR 50.2. The term “design bases” is defined in 10 CFR 50.2 as information which identifies the specific functions to be performed by a structure, system, or component of a facility, and the specific values or ranges of values chosen for controlling parameters as reference bounds for design. The definition further clarifies that design basis values may be (1) constraints derived from generally accepted “state-of-the-art” practices for achieving functional goals or (2) requirements based on analyses of the effects of postulated accidents for which an SSC must meet specified functional goals. Design bases are connected to safety assessments in two ways: (1) the performance capabilities of SSCs, as established through evaluations of DBEs and BDBEs, and (2) the design parameters for SSCs, which are derived from the operational context in which the function is to be performed, including considerations of natural phenomena and environmental factors. The NRC has determined that the existing definition of design bases in 10 CFR 50.2 provides sufficient flexibility to support the use of determinate, data-backed thresholds. In practice, the NRC has already applied determinate and data-backed thresholds for the selection of design bases attributes in several areas, including high winds, flooding, and seismic hazards. Therefore, a rulemaking to revise the definition of design bases is not necessary. However, the NRC is issuing draft guidance (DG)

The term BDBE has not previously been defined in 10 CFR 50.2. However, lessons learned from ongoing studies of nuclear plant risks, as well as operational experience, have historically led the NRC to identify and address plant events and conditions beyond the originally defined set of DBEs that could result in the release of radioactive material sufficient to pose a hazard to public health and safety. Accordingly, the NRC has imposed additional requirements to address such events when risk insights emerged from operational experience ( e.g., SBO in 10 CFR 50.63 and anticipated transients without scram (ATWS) in 10 CFR 50.62, “Requirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power plants”). These requirements extended regulatory attention beyond the traditional scope of DBEs. When developing this proposed rulemaking, the NRC initially considered a framework where these types of events were included in the DBE category. However, experience with regulating events such as SBO and ATWS has demonstrated that these types of events can be adequately addressed without the same regulatory treatment as DBEs. Therefore, the formal inclusion of the BDBE category in this proposed rule would provide a framework for applying graded regulatory treatment to such events. It would enable the NRC to address risks to public health and safety that do not warrant mitigation exclusively through safety-related SSCs or conservative safety assessments. Consequently, the NRC proposes to provide a definition for BDBE in 10 CFR part 50.

VI. Discussion—Determinate and Data-Backed Thresholds for Reactor Safety Assessments

The proposed changes would revise 10 CFR 50.2 to add definitions for the terms “design basis events” and “beyond design basis events.” The proposed changes would include a conforming revision to the definition of “design basis event” in 10 CFR 50.49(b)(1)(ii). These changes would apply to future 10 CFR part 50 and 52 applications submitted on or after the date that would be 180 days after the effective date of a final rule if this proposed rule were issued as a final rule; however, existing applicants, licensees, and approval holders under 10 CFR part 50 or 52 could voluntarily choose to adopt them. In parallel with the proposed changes, the NRC has developed DG-1454, which would (1) describe determinate and data-backed thresholds for categorizing events as DBEs or BDBEs, (2) outline graded assessment approaches for each event category, and (3) clarify the process for selecting design bases controlling parameters used as reference bounds in the design of SSCs. This guidance would use initiating event frequencies and qualitative criteria for categorizing events, maintaining consistency with the current safety assessment framework described in 10 CFR part 50 (and referenced in 10 CFR part 52). In addition, this guidance would describe acceptable approaches for identifying, grouping, and quantifying initiating events to ensure they are binned into appropriate categories. While alternate approaches, such as defining thresholds in terms of event sequences could be used, they would typically require the development of a full risk assessment or other systematic risk evaluation to determine sequence frequencies. To avoid imposing additional requirements not currently included in 10 CFR part 50, anchoring event selection to initiating event frequencies would provide a determinate and data-backed approach without adding regulatory burden.

Adding generally applicable definitions for DBE and BDBE would improve regulatory clarity and enable the use of objective criteria in selecting initiating events. Establishing threshold criteria, graded assessment approaches, and the selection process for design basis parameters within guidance would ensure that applicants and licensees are provided with an approach acceptable to the NRC, while ensuring flexibility for applicants to justify unique approaches, if desired, without the need for an exemption.

The addition of a definition of BDBE and corresponding thresholds would allow a reduction in unnecessary conservatism applied in the safety assessments of lower frequency events. The NRC concluded that reactor safety assessments associated with BDBEs are within the scope of the existing contents of application requirements of 10 CFR 50.34 and analogous sections of 10 CFR part 52. Specifically, requirements related to analysis and evaluation of the design and performance of SSCs of the facility with the objective of assessing the risk to public health and safety include consideration of BDBEs. Several existing regulations already address specific events not originally considered in the licensing basis or considered BDBEs (examples include but are not limited to ATWS, loss of all alternating current power events resulting in SBOs, and combustible gas control). For current applicants, licensees, or approval holders who may opt to adopt the proposed definitions for DBE and BDBE, as well as future applicants who would be mandated to use the proposed definitions, this rulemaking would not change the treatment of BDBEs specifically addressed by regulation such as ATWS and SBO. However, adoption of the BDBE definition could eliminate some events not specifically addressed by regulation from consideration that are determined to be non-credible.

Similarly, evaluations that assume substantial release of fission products would still be performed in accordance with 10 CFR 50.34(a) and 10 CFR 50.67, “Accident source term.” For LWRs, the release would be into containment. For other designs, it may be expressed as releases to the environment considering expected demonstrable leakage rates from potential flow paths and any fission product cleanup systems intended to mitigate the consequences of accidents. These evaluations would address the safety features that are engineered into a facility and those barriers that must be breached as a result of an accident before a radiological release to the environment can occur. Evaluations required to comply with 10 CFR 50.34(a) and 10 CFR 50.67 rely on conservative modeling assumptions. For example, as described in RG 1.183, “Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors,” source term fission product release fractions are derived from a set of accident sequences and many physical processes and phenomena are represented by bounding assumptions rather than being modeled directly. In addition, these evaluations credit only safety-related features in providing mitigation capability. Therefore, this evaluation is generally included in the spectrum of DBEs analyzed in Chapter 15 of NUREG-0800 ( e.g., Sections 15.0.1 or 15.0.3).

The current definition of design bases in 10 CFR 50.2 provides that controlling parameters may be derived either from accepted “state-of-the-art” practices or from analysis (based on calculations or experiments). The NRC has determined that this definition is sufficiently broad to accommodate the use of determinate, data-backed thresholds as implemented through guidance without the need for a rulemaking change. Design bases are identified through two primary means:

VII. Background—Removal of IEEE-323-1974 Reference in Footnote 3 of 10 CFR 50.49

Safety-related structures, systems and components are defined in 10 CFR 50.2, “Definitions.” The relationship between safety-related electric equipment and Class 1E equipment was initially established through footnote 3 of 10 CFR 50.49, “Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants.” The final rule promulgating 10 CFR 50.49, including footnote 3 of 10 CFR 50.49, (48 FR 2733; January 21, 1983) stated, in part: “The scope of the final rule covers that portion of equipment important to safety commonly referred to as “safety-related” (which the Commission interprets as essentially “Class 1E” equipment defined in [Institute of Electrical and Electronics Engineers (IEEE)]-323-1974).”

The connection between “safety-related” and “Class 1E” is now established in a more up-to-date standard—IEEE Standard 308, “IEEE Standard Criteria for Class 1E Power Systems for Nuclear Power Generating Stations,” which the NRC endorsed in RG 1.32, “Criteria for Power Systems for Nuclear Power Plants.”

VIII. Discussion—Removal of IEEE-323-1974 Reference in Footnote 3 of 10 CFR 50.49

The proposed action would remove footnote 3 of 10 CFR 50.49. Footnote 3 references an old standard that is no longer utilized as the sole means to establish the connection between “safety-related” and “Class 1E.” Instead, this connection is established in the more up-to-date IEEE Standard 308, which the NRC endorsed in RG 1.32. Removal of this footnote would improve regulatory clarity and would be consistent with the NRC modernizing and improving its regulations to reflect best practices and the maturity of the nuclear industry.

In addition, a minor editorial change is proposed to redesignate the current footnote 4 of 10 CFR 50.49 as footnote 1 given the previous and proposed deletions of the preceding footnotes.

IX. Background—Expanded Alternative Requests Under 10 CFR 50.55a(z)

In 10 CFR 50.55a, “Codes and standards,” the NRC incorporates by reference certain parts of editions and addenda of specified codes and standards through rulemaking. Upon incorporation by reference of these specified codes and standards into 10 CFR 50.55a, the provisions of these codes and standards are legally-binding NRC requirements as delineated in 10 CFR 50.55a, subject to the conditions on certain specific provisions that are set forth in 10 CFR 50.55a. Currently, in paragraph (z), “Alternatives to codes and standards requirements,” of 10 CFR 50.55a, an applicant or licensee may request authorization of alternatives to the requirements of paragraphs (b), “Use and conditions on the use of standards,” through (h), “Protection and safety systems,” of 10 CFR 50.55a, if the applicant or licensee demonstrates either that the proposed alternative would provide an acceptable level of quality and safety or that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Since its initial promulgation in 1971 (36 FR 11423; June 12, 1971), 10 CFR 50.55a has allowed for the consideration of proposed alternatives under these same two criteria. Over the years, 10 CFR 50.55a has been periodically updated to reflect revised and updated codes and standards for nuclear power plants.

On March 15, 1984 (49 FR 9711), the NRC issued a final rule that made procedural changes by, among other things, clarifying the procedures for alternatives, expressly noting that alternatives can be authorized by the Director of the Office of Nuclear Reactor Regulation. In the November 5, 2014, final rule, “Approval of American Society of Mechanical Engineers' [ASME] Code Cases” (79 FR 65776), the NRC restructured 10 CFR 50.55a to align with the Office of the Federal Register's guidelines for incorporation by reference and to allow proposed alternatives to NRC-approved Code Cases rather than only to ASME Code provisions. In this restructuring, the proposed alternatives provisions were moved from their prior location in paragraph (a)(3) of 10 CFR 50.55a to a newly designated paragraph (z) of 10 CFR 50.55a. However, these rulemakings addressed only procedural clarifications and a restructuring of existing regulations, not changes in the scope of opportunities for alternatives. In the July 17, 2024, final rule, “American Society of Mechanical Engineers Code Cases and Update Frequency” (89 FR 58039), the NRC added paragraph (y), “Definitions,” to 10 CFR 50.55a. These definitions provide consistency and clarity throughout 10 CFR 50.55a and accommodate new opportunities to change code of record intervals. However, these definitions were added outside the scope of paragraph (z) of 10 CFR 50.55a because the Commission had not approved the use of 10 CFR 50.55a(z) for definitions or the newly defined intervals.

X. Discussion—Expanded Alternative Requests Under 10 CFR 50.55a(z)

Currently, the proposed alternative provisions of paragraph (z) of 10 CFR 50.55a apply to the codes and standards requirements in paragraphs (b) through (h) of 10 CFR 50.55a. The addition of paragraph (y) to 10 CFR 50.55a in the July 17, 2024, final rule, without an associated expansion of the scope of paragraph (z), has resulted in unanticipated exemptions under 10 CFR 50.12, “Specific exemptions,” to use alternate definitions to those included in paragraph (y) of 10 CFR 50.55a. Moreover, the criteria of paragraph (z)(1), “Acceptable level of quality and safety,” or (z)(2), “Hardship without a compensating increase in quality and safety,” of 10 CFR 50.55a provide appropriate controls for requested alternatives to all requirements in 10 CFR 50.55a, so there is no need to restrict the application of 10 CFR 50.55a(z) to only some paragraphs in 10 CFR 50.55a. Therefore, the NRC proposes to remove the restriction limiting proposed alternatives to paragraphs (b) through (h) of 10 CFR 50.55a so that proposed alternatives

In addition to creating paragraph (y) of 10 CFR 50.55a, the revisions in the July 17, 2024, final rule provided more flexibility to licensees by expanding the code of record interval from 10 years to two consecutive inservice testing and inservice inspection intervals. In that final rule's preamble, the Commission stated, in part, that licensees may request future alternatives based upon the code of record interval. This revision, coupled with the staff position in SECY-23-0061, “Clarification of the Staff's Position on Certain American Society of Mechanical Engineers Code Alternatives for More Than One 10-Year Inservice Inspection Interval Under Title 10 of the Code of Federal Regulations 50.55a,” dated July 21, 2023, clarified that, when appropriately justified, the duration of an alternative need not be limited to the length of a single inservice testing or inservice inspection interval. Rather, the NRC may approve specific alternatives for longer durations when the technical bases supporting the requested alternative ensure that an acceptable level of quality and safety will be maintained. These existing flexibilities would be unchanged by this proposed rule.

The proposed change would provide additional flexibility to a licensee or applicant, while maintaining the same requirements for an acceptable level of quality and safety or the presence of a hardship without a compensating increase in quality and safety, which have been foundational to proposed alternatives since their initial promulgation.

XI. Background—Risk-Informing 10 CFR 50.59 and Allowing Flexibility for Changes to Methods

A. Use of Quantitative Risk Results

The AEA requires a licensee to seek an amendment for significant changes to its facility or procedures. The Commission possesses substantial discretion to define, by rule, the threshold that constitutes a change significant enough to require a license amendment.

The current regulation in 10 CFR 50.59 was developed in response to issues involving inconsistency in how licensees applied the previous criteria to determine whether changes, tests, or experiments require prior NRC approval. The NRC is proposing to amend 10 CFR 50.59 to allow licensees to consider risk insights from PRAs when applying the criteria in that provision. The statements of consideration for the 10 CFR 50.59 final rule, “Changes, Tests, and Experiments” (64 FR 53582; October 4, 1999), did not allow licensees to use PRA insights at that time, but the Commission recognized the possibility that the NRC could one day develop the regulatory infrastructure to support the use of PRAs in 10 CFR 50.59 analyses.

The NRC now proposes to incorporate the use of quantitative risk results, like Core Damage Frequency (CDF) and Large Early Release Frequency (LERF), to evaluate changes under 10 CFR 50.59(c)(2)(i) and (ii).

The current state of practice in the nuclear fleet for quantitative risk assessment is the use of Level 1/limited Level 2 PRAs. The quantitative risk metrics output by these PRAs are CDF and LERF, and they, along with their changes ( i.e., ΔCDF and ΔLERF), are the metrics used as part of risk-informed decision-making processes. As described in RG 1.174, Revision 3, “An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,” dated January 2018, these risk metrics are based on the Commission's safety goals and the associated quantitative health objectives. Similarly, the NRC proposes to use the small changes in CDF and LERF jointly as means to determine the importance of the effect of the proposed change on accident frequency and SSC malfunction rate under 10 CFR 50.59.

B. Improved Flexibility for Changes to Methods of Evaluation

Currently, 10 CFR 50.59 allows licensees to make certain changes to their facility or procedures without prior NRC approval, provided those changes do not meet specific thresholds that would require a license amendment. One of those thresholds, stated in 10 CFR 50.59(c)(2)(viii), requires NRC review of any change in a methodology that results in a departure from a method of evaluation described in the FSAR (as updated) used in establishing the design bases or in the safety analysis. As explained in the preamble of the 10 CFR 50.59 final rule in 1999, this language was chosen to ensure NRC oversight of the safety margins and conservatisms that form the basis of the NRC's licensing decision. The Commission stated that the language of 10 CFR 50.59(c)(2)(viii) was selected “to allow licensees only a small degree of flexibility in methods where the results are tending in the non-conservative direction” (64 FR 53598; October 4, 1999).

In SECY-97-035, “Proposed Regulatory Guidance Related to Implementation of 10 CFR 50.59 (Changes, Tests, and Experiments),” dated February 12, 1997, which transmitted proposed 10 CFR 50.59 guidance to the Commission ahead of the proposed rulemaking, the staff recognized that, “as the knowledge base increases and computing power increases, new methods of analysis will more accurately predict the actual plant response.” However, the staff found that a comparison of the analytical results from two different methodologies was not valid to make a 10 CFR 50.59 determination. To make the 10 CFR 50.59 determination using a new methodology, the new methodology must be valid ( e.g., previously approved by the NRC) and the analysis in question must be performed for the situation before the change and the situation after the change using the same methodology.

In the decades since the original rule was written, there have been substantial advancements in computational capabilities and modeling practices. New data, improved understanding, and increased computing power now allow for faster iteration and refinement of methods used in safety analyses. Due to this substantial increase in computational power and the rapid growth in the use of modeling and simulation, other industries are beginning to shift away from model-by-model reviews and are instead focusing on the processes by which organizations establish the credibility of their models—specifically through verification, validation, and uncertainty quantification (VVUQ) programs. The U.S. Food and Drug Administration has taken the most prominent step in this direction by issuing new guidance centered on credibility assessments (FDA-2021-D-0980; November 17, 2023). The aviation industry is actively developing a VVUQ standard through an industry-led, Federal Aviation Administration-supported initiative.

The existing requirements of 10 CFR 50.59(c)(2)(viii) mandate a license amendment for any departure from a method of evaluation described in the FSAR. As defined in the regulation, a departure is a change to any element of a methodology, unless the results are

XII. Discussion—Risk-Informing 10 CFR 50.59 and Allowing Flexibility for Changes to Methods

A. Use of Quantitative Risk Results

The NRC proposes rulemaking to establish an alternative pathway that would allow the use of quantitative risk metrics, like CDF and LERF, along with consideration of safety margins and defense in depth, to evaluate a proposed change, test, or experiment against the criteria of 10 CFR 50.59(c)(2)(i) and (ii), while leaving all other criteria in 10 CFR 50.59(c)(2) in place. New proposed 10 CFR 50.59(e) would establish a risk-informed alternative to the existing regulation and would not alter or impede the current practice for evaluating proposed changes against the text of 10 CFR 50.59(c)(2)(i) and (ii), as written. A licensee could continue to use qualitative assessments, engineering judgement, and other existing practices and techniques to evaluate a proposed change, test, or experiment.

Under the proposed 10 CFR 50.59(e), a licensee could demonstrate that a change would not result in a “more than a minimal increase” under 10 CFR 50.59(c)(2)(i) and (ii) by using quantitative risk results based on a PRA of appropriate scope and quality that provides appropriate risk metrics. The change would also need to maintain defense-in-depth and safety margins. “Appropriate scope and quality” in this context would mean that the licensee's model fully encompasses the proposed change and that the model has been found to be acceptable for use in a previous NRC-approved application. “Appropriate risk metrics” in this context would mean quantitative results that demonstrate the effects on the proposed change and can provide a baseline for judging facility risk. For traditional PRAs, these metrics are CDF, LERF, and the changes (Δs) to CDF and LERF. Extensive discussion of maintaining defense in depth and safety margins can be found in RG 1.174, Revision 3. The NRC has proposed guidance for 10 CFR 50.59(e) in DG-1466, draft Revision 4 to RG 1.187, “Guidance for Implementation of 10CFR50.59, `Changes, Tests, And Experiments.'”

The use of PRA would not replace or supplant the deterministic licensing basis but would supplement it with a powerful analytical tool. Since the initial licensing of the current fleet, the NRC and the industry have developed and matured PRA methodologies, which provide a holistic, integrated assessment of plant safety. PRA can identify contributors to risk and potential vulnerabilities that may not be apparent from a purely deterministic analysis.

Incorporating CDF and LERF into the 10 CFR 50.59 process would not be an attempt to re-license plants on a probabilistic basis. Instead, it would provide an alternative methodology licensees could voluntarily choose for conducting analyses under 10 CFR 50.59. It would use risk insights to inform the judgment of the safety significance of changes to the existing deterministic design. A change that results in a very small, quantifiable increase in calculated risk could be reasonably judged not to undermine the fundamental safety basis established through deterministic principles. This approach would allow for a more consistent, predictable, and efficient screening process, directly fulfilling the original purpose of 10 CFR 50.59 to differentiate between changes that require prior NRC review and those that do not.

This integration would be consistent with decades of evolving NRC policy and practice. The agency has successfully used risk-informed approaches in many other regulatory applications, including 10 CFR 50.65, “Requirements for monitoring the effectiveness of maintenance at nuclear power plants”; 10 CFR 50.48(c), “National Fire Protection Association Standard NFPA 805”; the Reactor Oversight Process; and RG 1.174. Using quantitative risk results, such as from a PRA, in the 10 CFR 50.59 process would be a logical evolution that would enhance the existing framework by leveraging modern analytical tools to better focus licensee and agency resources on issues of genuine safety significance.

B. Improved Flexibility for Changes to Methods of Evaluation

The NRC proposes two regulatory amendments, which would work in concert to allow licensees greater flexibility to implement changes to analytical methods described in the FSAR.

First, the NRC proposes a targeted revision to 10 CFR 50.59(c)(2)(viii) and 53.1550(a)(2)(iv). This change would allow licensees to implement certain changes to analytical methods described in the FSAR (as updated) without prior NRC approval, provided those changes are undertaken pursuant to an NRC-approved VVUQ program under 10 CFR 50.221, “Credibility requirements for modeling and simulation.” Proposed guidance for compliance with proposed 10 CFR 50.221 is in DG-1468, “Guidance for Implementation of 10 CFR 50.221, `Credibility requirements for modeling and simulation.'” This revision would clarify that appropriate changes made using a risk-informed and graded VVUQ framework would be permissible under 10 CFR 50.59 without prior NRC approval. The proposed rule would allow such changes without prior NRC approval only if the VVUQ program were approved by the NRC for the method of evaluation in question, and the new method of evaluation met the credibility criteria established in the approved VVUQ. These measures would maintain safety and ensure appropriate controls over licensee changes to methods or evaluations while affording flexibility through reliance on the NRC-approved VVUQ program.

Second, the NRC proposes to adopt an optional regulation on VVUQ at 10 CFR 50.221 that would establish the requirements a VVUQ program must meet. The new regulation would establish clear requirements and structure for VVUQ activities used to support regulatory decisions. Specifically, the proposed 10 CFR

Via these proposed changes, the NRC would shift focus away from evaluating each individual model or simulation directly and toward evaluating the process by which models are determined to be credible—specifically through structured VVUQ programs. In this context, credibility would refer to the level of trust in a model's ability to produce accurate and appropriate predictions for its intended use. An NRC-approved VVUQ program would establish a new licensing basis that focuses on how methods are selected and applied, rather than the characteristics of the specific method and the inherent conservatisms. Where appropriate, VVUQ-based credibility assessments could serve as a viable alternative to full NRC review of each new or revised model. However, a single generic VVUQ process would not be appropriate due to the significant variability in physical phenomena, modeling assumptions, numerical techniques, and uncertainties across different reactor technologies and methods of evaluation. A tailored VVUQ process approved by the NRC for the intended application would be required for each method of evaluation to provide the necessary specificity to ensure credible, defensible assessments of model performance for each unique application for all possible reactor technologies and their vastly different physical domains.

This approach could be particularly beneficial for new and advanced reactor designs. Unlike the current fleet, which has decades of operational data and analytical stability, advanced reactors often lack extensive experimental databases at the time of initial licensing. Requiring them to demonstrate method maturity at the level of existing plants would demand significant upfront testing and analysis, delaying deployment and increasing cost. Many of these designs are being developed as test reactors specifically to generate such data. A rule change would provide a clear, structured mechanism for these reactors to update their methods over time, based on data collected during operation, without needing to go through repeated full NRC reviews—so long as the updates are made through an approved VVUQ process.

XIII. Background—Minimum Decommissioning Funding Assurance Requirements for Non-Large Light-Water Reactors

The regulation in 10 CFR 50.75, “Reporting and recordkeeping for decommissioning planning,” establishes requirements for indicating to the NRC how an applicant or licensee will provide reasonable assurance that funds will be available for the decommissioning process. During the operational phase of a reactor facility, an applicant or licensee must certify that funding is being provided in an amount that may be more, but not less, than the amount described in 10 CFR 50.75(c)(1) and (2) (also known as the table of minimum amounts, minimum funding assurance, or “formula” amount). As the NRC stated in the 1988 decommissioning rule (53 FR 24018-24030; June 27, 1988), the “formula” amount in 10 CFR 50.75(c) does not represent the actual cost of decommissioning for specific reactors but rather serves as a reference level established to ensure that the bulk of the funds necessary for a safe decommissioning is being considered and planned for early in facility life by licensees. This provides assurance that the facility will not become a risk to public health and safety when it is decommissioned.

The table of minimum amounts and the associated adjustment factors were developed and designed specifically for the large light-water reactor technologies (boiling water reactors and pressurized water reactors) that make up the current commercial power reactor fleet in the U.S. However, new reactors may incorporate different technologies and output capacities that may not require the amount of decommissioning funding assurance described in 10 CFR 50.75(c). In order to address different decommissioning funding needs for these new technologies without requiring an exemption from NRC regulations, the NRC is proposing updates to its regulations to allow certain new reactor applicants and licensees the flexibility to certify adequate decommissioning funding assurance during operations through the use of either the table of minimum amounts or the submission of a design-specific decommissioning cost estimate. Allowing for the use of a design-specific decommissioning cost estimate that may be less than the table of minimum amounts would provide a path for certain new reactor applicants and licensees to demonstrate financial responsibility for safe decommissioning based on factors specific to the reactor facility.

XIV. Discussion—Minimum Decommissioning Funding Assurance Requirements for Non-Large Light-Water Reactors

The NRC proposes an amendment to 10 CFR 50.75 to allow certain new reactor applicants and licensees to submit a design-specific decommissioning cost estimate to demonstrate minimum decommissioning funding assurance during operations that may be less than the table of minimum amounts provided in 10 CFR 50.75(c). The values in the current table of minimum amounts are based on funding assumptions associated with the decommissioning of large light-water reactor facilities. This proposed rule would allow certain new reactor applicants and licensees to certify financial assurance for decommissioning through the use of either the minimum formula amount or through the submission of a design-specific decommissioning cost estimate. Thus, the proposed rule would provide flexibility for new reactor applications that represent smaller output and size considerations than large light-water reactor designs.

Specifically, the NRC is proposing to add a new paragraph (b)(2) to 10 CFR 50.75 that describes the certification amount process and minimum requirements, including reliance on design-specific decommissioning cost estimates, for new reactor applicants and licensees seeking to use the alternative pathway. Additionally, the NRC is proposing to delete language in the table of minimum amounts in 10 CFR 50.75(c)(1) that requires reactors of less than 1200 megawatts thermal (MWt) to use the certification amount for a 1200 MWt reactor. It is conceivable that new LWR designs could have an output less than 1200 MWt. Therefore, requiring a new reactor applicant or

A certification relying on a design-specific decommissioning cost estimate would be required to include a description of the factors used to develop the design-specific decommissioning cost estimate, including generic activities performed in the major decommissioning phases of a decommissioning project ( e.g., pre-decommissioning engineering and planning, reactor deactivation, and dismantlement). Similar to the table of minimum amounts for large light-water reactor designs, design-specific decommissioning cost estimates should represent the bulk of funds necessary to safely decommission a facility, as applicable to the specific reactor technology being utilized and the design of the facility. Additionally, similar to the table of minimum amounts for large light-water reactor designs, the design-specific decommissioning cost estimate would have to be adjusted annually at a rate at least equal to the formula in 10 CFR 50.75(c)(2). This certification process would include NRC review and approval. However, once an initial design-specific decommissioning cost estimate is approved by the agency as a sufficient certification amount for financial assurance for decommissioning, other applicants or licensees using similar technology could reference and justify use of this amount (escalated in accordance with NRC regulations and guidance) as the certification amount required by proposed 10 CFR 50.75(b) for a different application. Once a licensee nears permanent cessation of operations, a site-specific decommissioning cost estimate that encompasses the design-specific cost as well as costs associated with the site and operational period of the facility, would be required for funding assurance purposes, as described in current regulations in 10 CFR 50.82, “Termination of license.” Finally, the NRC is proposing to revise 10 CFR 50.75(e)(1)(i) and (ii) to allow reactor licensees that have prepaid or collected funds based on a design-specific estimate to take credit for projected earnings on the prepaid or collected decommissioning funds using up to a 2-percent annual real rate of return up to the time of permanent termination of operations.

The NRC is proposing a similar change to 10 CFR part 53. Currently, 10 CFR part 53 only allows for a site-specific decommissioning cost estimate as the certification amount. Therefore, the NRC is proposing to add conforming language to 10 CFR 53.1010, “Financial assurance for decommissioning,” and 53.1020, “Cost estimates for decommissioning,” to allow new reactor applicants and licensees to submit a design-specific decommissioning cost estimate to demonstrate minimum decommissioning funding assurance during operations. Finally, the NRC is proposing to add conforming language to 10 CFR 53.1040, “Methods for providing financial assurance for decommissioning,” to allow new reactor applicants and licensees that have prepaid or collected funds based on a design-specific estimate to take credit for projected earnings on the prepaid or collected decommissioning funds using up to a 2-percent annual real rate of return up to the time of permanent termination of operations.

Additionally, conforming changes would be made to 10 CFR 50.75(e) to include references to the requirements of proposed 10 CFR 50.75(b)(2), where appropriate. In addition, this proposed rule would revise 10 CFR 50.75(e) to include “applicant or” in all appropriate places where currently only “licensee” is referenced, as directed by the Commission in staff requirements memorandum (SRM)-SECY-23-0021: Enclosure 4, “Table of Typographical errors and Inconsistencies,” dated March 4, 2024. Similar changes are proposed in 10 CFR 53.1040 and 10 CFR 53.1050, “NRC oversight” for consistency. Collectively, these changes would clarify that applicants and licensees would be subject to the requirements under 10 CFR 50.75(e), 53.1040, and 53.1050, as applicable.

This proposed rule also would make minor editorial changes in 10 CFR 50.75(e), (g), and (h) by removing errant commas, correcting capitalization errors, correcting references by indicating paragraphs instead of sections, and removing “of this part,” where necessary.

XV. Background—Incorporation of Streamlined Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants

A. Historic Quality Assurance Requirements Perspectives and Emergent Issues

During the early days of nuclear power (1950s-1960s), the Atomic Energy Commission (AEC), the NRC's predecessor agency, focused on developing and licensing nuclear reactors. As the nuclear industry grew, it became clear that systematic quality assurance (QA) was essential to ensure nuclear safety, especially given the complexity and potential hazards of nuclear technology. By the 1960s, nuclear power plants were becoming more complex, and the consequences of failures were more severe. Incidents and near-misses highlighted the need for formalized QA programs to prevent design, fabrication, and construction errors. Appendix B was added to 10 CFR part 50 by the AEC in 1970 (35 FR 10498; June 27, 1970) to (1) establish minimum QA requirements for safety-related structures, systems, and components (SSCs) and (2) ensure that these SSCs are designed, fabricated, constructed, and tested to perform their intended safety functions.

Applicants for CPs, OLs, early site permits (ESPs), COLs, design certifications, standard design approvals, and manufacturing licenses (MLs) must include in their respective application a description of the QA program that discusses how the applicable requirements of appendix B to 10 CFR part 50 are satisfied.

Although appendix B to 10 CFR part 50 is foundational to nuclear safety, commenters have expressed concerns over its implementation, flexibility, and alignment with modern practices. These concerns include the following topics.

• Inflexibility: Appendix B to 10 CFR part 50 (1) is prescriptive and has not been substantively updated since 1970; (2) lacks risk-informed or performance-based flexibility, which modern quality systems increasingly emphasize; and (3) does not facilitate tailoring of QA programs to low-risk activities or innovative technologies.

• Outdated Language: The language in appendix B to 10 CFR part 50 is reflective of technologies from the 1970s and does not explicitly address digital systems, software QA, and additive manufacturing. As a result, applicants must rely on guidance, which could result in inconsistent implementation across applicants due to applicants interpreting the guidance differently based on their specific technologies.

• Vendor and Supply Chain Challenges: Many suppliers, especially non-nuclear vendors, are unfamiliar with appendix B to 10 CFR part 50, and thus applicants and licensees have challenges in procuring products and services for the nuclear power plants.

• Lack of Harmonization with International Standards: Appendix B to 10 CFR part 50 is United States-specific and not aligned with international standards and best practices and, thus, creates challenges for international collaboration and global supply chains.

In light of these considerations, commenters have advocated for a modernized, risk-informed QA framework that retains the safety rigor of appendix B to 10 CFR part 50 while allowing for graded application based on safety significance that is integrated with modern quality systems.

B. NRC Responses to These Issues

As a result of these issues, the NRC is proposing to add appendix T to 10 CFR part 50 as a voluntary alternative to appendix B to 10 CFR part 50. The proposed appendix T would draw on international QA standards to incorporate the following elements:

• Performance-based QA criteria that provide explicit direction on use of a graded approach for applying QA requirements to SSCs relative to their safety and risk contributions to the overall nuclear facility.

• Quality assurance requirements specific to software used in digital items and for design and analysis.

• Quality assurance terminology and methodologies used across various safety critical industries and in international standards for quality management, thus allowing applicants to leverage cross-industry and global supply chains.

XVI. Discussion—Incorporation of Streamlined Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants

This proposed rule would add a new appendix T to 10 CFR part 50 to provide streamlined QA criteria that could be used for applications of COLs, CPs, and OLs under certain eligibility requirements.

A. Introduction and Scope

As discussed in section XV, “Background—Incorporation of Streamlined Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,” of this document, NRC stakeholders have expressed interest in utilizing a streamlined approach to QA that better aligns with international standards. Therefore, the NRC has developed a proposed appendix T to 10 CFR part 50 that would provide an alternative to the current QA requirements in appendix B to 10 CFR part 50 based on International Standard, ISO 19443, “Quality management systems—Specific requirements for the application of ISO 9001:2015 by organizations in the supply chain of the nuclear energy sector supplying products and services important to nuclear safety [ITNS],” (2018-05). The International Standard Organization collaborated closely with the International Atomic Energy Agency in developing ISO 19443.

The proposed section I, “Introduction and Scope,” of appendix T to 10 CFR part 50 would provide the eligibility requirements for using appendix T to 10 CFR part 50 as a voluntary alternative to appendix B to 10 CFR part 50. Specifically, applicants for CPs, OLs, and COLs would have the option to use appendix T to 10 CFR part 50 as an alternative to appendix B to 10 CFR part 50, provided that the three conditions in section I are met.

Proposed condition I.A would require that the application is for an nth-of-a-kind (NOAK) plant and would require the application to identify the first-of-a-kind (FOAK) reference plant.

Proposed condition I.B would require that any departures from the FOAK reference plant in the application of the NOAK plant would not result in a change to the classification, design, and method of manufacture, construction, and operation of SSCs identified in licensing basis of the referenced plant.

Proposed condition I.C would require the application to include procedures and work processes for implementing the requirements in proposed appendix T to 10 CFR part 50.

The NRC would define FOAK nuclear power plants and fuel reprocessing plants in proposed appendix T as the initial implementation of a new reactor design or technology or new fuel reprocessing plant design that has not been previously constructed and operated at commercial scale, either within the U.S. or internationally. The FOAK plant would serve as a reference plant for NOAK nuclear power plants and fuel reprocessing plants. The NRC would define NOAK nuclear power plants and fuel reprocessing plants in proposed appendix T to be any subsequent implementation of a FOAK reactor design or technology or fuel reprocessing plant design after the FOAK plant has been designed, constructed, and operated.

Proposed conditions I.A, I.B, and I.C would limit the use of proposed appendix T to 10 CFR part 50 to those applications that could potentially leverage the following:

• the design maturity of the FOAK completed plant design information;

• standardized components and systems to streamline procurement and construction;

• skilled labor and contractors from the FOAK projects;

• FOAK operational data to help NOAK commissioning procedures, standard operating procedures, emergency operating procedures, training, design, and operational programs; and

• established processes and procedures for implementing the requirements in proposed appendix T to 10 CFR part 50, including an established mechanism to ensure deviations from FOAK reliability in NOAK would be identified and promptly corrected.

In recent cases, the NRC has noticed that design details and QA procedures are not available during the licensing review of a CP, OL, or COL for a FOAK plant. Therefore, during licensing and construction for FOAK plants, the NRC conducts vendor inspections to confirm the detailed design and as-built SSCs meet the technical and quality requirements committed to by the applicant. However, because a NOAK applicant could reference design details and QA procedures developed during the FOAK licensing, these additional QA activities, such as vendor inspections, may be unnecessary, especially for vendors previously inspected by the NRC. Therefore, when the conditions I.A, I.B, and I.C are met, the proposed appendix T to 10 CFR part 50 may eliminate the need for NRC oversight of suppliers and vendors who supply applicants with approved appendix T to 10 CFR part 50 compliant QA programs because (1) the list of SSCs that would be governed by proposed appendix T to 10 CFR part 50 would be standardized and verified to be acceptable using information from the reference FOAK plant; (2) the design of these SSCs would be complete and verified to be acceptable using information from the reference FOAK plant; and (3) the manufacturing and construction methods for these SSCs would be established and verified to be acceptable during oversight of manufacturing and construction processes for the referenced FOAK plant.

B. Definitions

The proposed section II, “Definitions,” of proposed appendix T to 10 CFR part 50, would include definitions for terms used in the proposed appendix.

The NRC would define “first-of-a-kind” nuclear power plants and fuel reprocessing plants as the initial implementation of a new reactor design or technology or new fuel reprocessing plant design that has not been previously constructed and operated at commercial scale.

The NRC would define “nth-of-a-kind” nuclear power plants and fuel reprocessing plants as any subsequent implementation of a reactor design or technology or fuel reprocessing plant

The NRC would define “quality assurance” as all those planned and systematic actions necessary to provide adequate confidence that a structure, system, or component will perform satisfactorily in service. Quality assurance includes quality control, which comprises those actions related to the physical characteristics of a material, structure, component, or system that provide a means to ensure the material, structure, component, or system meets predetermined requirements. This proposed definition is equivalent to the definition used in appendix B to 10 CFR part 50.

The NRC would define “quality assurance program” as the overall program established to assign responsibilities and authorities, define policies and requirements, and provide for the performance and assessment of work necessary to achieve QA.

The NRC would define “quality management system” (QMS) as a structured framework that documents an organization's processes, procedures, and responsibilities for ensuring quality. This term and definition are used broadly by other safety-critical industries, nuclear regulatory bodies and industry abroad, and vendors and third-party suppliers to these industries. The QMS is different from the terminology “quality assurance program description” (QAPD) used in appendix B to 10 CFR part 50 in that the QMS has a broader scope and is a system framework that includes quality planning, controls, assurance, and improvement; whereas a QAPD is a descriptive document that is narrowly focused on QA.

The NRC would define “functional design criteria” as metrics for the performance of SSCs. For safety-related SSCs, these criteria define performance metrics necessary to demonstrate compliance with the safety criteria in 10 CFR 53.210, “Safety criteria for design-basis accidents.” For non-safety-related but safety-significant SSCs, these criteria define performance metrics necessary to demonstrate compliance with the safety criteria in 10 CFR 53.220, “Safety criteria for licensing-basis events other than design-basis accidents.” This proposed definition would be added to proposed appendix T to 10 CFR part 50 to align with the definition and use of this term in 10 CFR part 53.

The NRC would define “non-safety-related but safety-significant SSCs” as those SSCs that are not safety-related but are relied on to achieve adequate defense in depth or perform risk-significant functions and warrant special treatment. This proposed definition would be added to proposed appendix T to 10 CFR part 50 to align with the definition and use of this term in 10 CFR part 53.

C. General Requirements

The proposed section III, “General Requirements,” of proposed appendix T to 10 CFR part 50 would provide general requirements for establishing and maintaining a QA program for applicants that choose to meet proposed appendix T.

Proposed section III.A, “Integrated Quality Assurance Program,” of proposed appendix T to 10 CFR part 50 would require that the integrated QA program ensures that safety-related and non-safety-related but safety-significant SSCs are designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions those SSCs perform. Proposed section III.A of in proposed appendix T to 10 CFR part 50 would include the following seven items that any application using in proposed appendix T to 10 CFR part 50 would be required to identify and explain in the integrated QA program:

• Responsibilities (1) are properly assigned to specific individuals or teams in charge of executing QA activities and (2) ensure any delegated responsibilities are properly identified and controlled.

• The design requirement of SSCs are sufficiently captured in corresponding documents; the design bases requirements are adequately translated into specifications, drawings, procedures, and instructions; outputs reflect the correct design inputs; the design is properly verified and validated; the as-built and as-operated SSC properly meet the intended function and safety margin.

• Means and methods are established to communicate relevant technical, quality, and regulatory requirements, expectations, and concerns between the applicant and its vendors and third-party suppliers.

• Measures are established to (1) ensure that procured SSCs and related services meet technical and quality requirements and (2) assess the capability of vendors or third-party suppliers that supply the SSCs and related services.

• Measures are established to (1) verify and validate that products and services meet the technical and quality requirements of the procured products and services, and (2) audit the vendors or third-party suppliers that are providing the products and services.

• Processes are implemented to address reoccurrence of issues and failures.

• Recordkeeping and documentation protocols for the QA program are established.

Proposed section III.B.1 of proposed appendix T to 10 CFR part 50 would require the applicant document the QA program in the QMS and submit the QMS to the NRC for review and approval. Proposed section III.B.1 would require the QA program, as documented in the QMS, to contain a graded approach for implementing the requirements of the QA program.

Proposed section III.B.2 of proposed appendix T to 10 CFR part 50 would require the QMS to describe how the requirements in section IV of the appendix would be met. Proposed section IV of appendix T to 10 CFR part 50 would identify general QA criteria and software QA criteria. Proposed section III.B.3 of appendix T to 10 CFR part 50 would require the applicant to invoke the applicant's QA requirements in procurement documents to all relevant contractors, vendors, suppliers, and third-parties.

Proposed section III.B.4 of proposed appendix T to 10 CFR part 50 would require the applicant to select the appropriate industry standards that are used to achieve quality consistent with regulatory requirements and the NRC's policies. This proposed section would state that the applicant would need to document the selection of ASME Nuclear Quality Assurance (NQA)-1, “Quality Assurance Requirements for Nuclear Facility Applications” or another appropriate industry standard. Proposed section III.B.4 of appendix T to 10 CFR part 50 would also require that gaps between the selected industry standards and the proposed section IV, “Quality Assurance Requirements,” of appendix T to 10 CFR part 50 are addressed within the QMS.

D. Quality Assurance Requirements

Proposed section IV.A, “Quality Assurance Criteria,” of proposed appendix T to 10 CFR part 50, identifies QA criteria that would be applicable to all applications that reference proposed appendix T to 10 CFR part 50. This proposed section would include 11 criteria that cover topical areas in management, performance, and assessment. The proposed requirements in these topical areas are consistent with International Standards for QMSs such as ISO 9001, “Quality Management System—Requirements,” which are used by many safety-critical industries, and ASME NQA-1, “Quality Assurance Requirements for Nuclear Facility

(i) Management

Proposed section IV.A.1, “Criterion 1—Management: Program,” of proposed appendix T to 10 CFR part 50, would include requirements for the applicant to define the organizational structure, functional responsibilities, levels of authority, and interfaces for performing the work necessary to implement the QA program, and develop management processes to plan, schedule, and assign resources to perform this work.

Proposed section IV.A.2, “Criterion 2—Management: Personnel Training and Qualifications,” of proposed appendix T to 10 CFR part 50, would include requirements for the applicant to develop processes for indoctrination and continuous training of employees for performing the work necessary to implement the QA program.

Proposed section IV.A.3, “Criterion 3—Management: Quality Improvement,” of proposed appendix T to 10 CFR part 50, would include requirements for the applicant to establish and implement a process for identifying and controlling issues and failures that could adversely impact quality, safety, and regulatory compliance. The proposed section IV.A.3 would also require applicants to include prevention of recurrence of issues as part of corrective actions and implement processes for continuous improvement of the QA program.

Proposed section IV.A.4, “Criterion 4—Management: Documents and the Associated Records,” of proposed appendix T to 10 CFR part 50, would include requirements for the applicant to prepare, review, approve, issue, use, and revise documents that prescribe processes, specify requirements, or establish the design of the SSC, and maintain these documents as records for the QA program.

(ii) Performance

Proposed section IV.A.5, “Criterion 5—Performance: Work Processes,” of proposed appendix T to 10 CFR part 50, would include requirements for the applicant to perform work, including hazard controls. Hazard controls are systematic measures designed to prevent, detect, and correct issues that could compromise quality, safety, and regulatory compliance. Examples of hazard controls applicable to QA programs include:

• embedded work processes to ensure quality such as inspection and testing protocols, hold points and witness points, and nonconformance reporting protocols;

• monitoring and detection programs to ensure detection of deviations and malfunctions in real time such as surveillance and audits;

• structured approaches to identify quality issues and implement corrective and preventive actions.

Proposed section IV.A.5 of proposed appendix T to 10 CFR part 50, would also require the applicant to establish and implement work processes for identifying and controlling items to ensure proper use; maintain items to prevent damage, loss, or deterioration; and calibrate and maintain equipment used for activities affecting quality.

Proposed section IV.A.6, “Criterion 6—Performance: Design,” of proposed appendix T to 10 CFR part 50, would include requirements for the applicant to establish and implement measures for controlling the design of SSCs, including requirements for controlling design changes, design interfaces, and verifying and validating the adequacy of the design. Verify in the context of design control means to perform the set of activities to demonstrate that design conforms to specifications and occurs during the design and development process. Examples include reviews, inspections, and unit tests. Validate in the context of design control means to perform the set of activities to demonstrate that the as-developed or as-built SSC performs the intended safety-functions and occurs after the development process. Examples include integrated tests, analysis, and simulations.

Proposed section IV.A.7, “Criterion 7—Performance: Procurement,” of proposed appendix T to 10 CFR part 50, would include requirements for the applicant to establish and implement processes for procurement of items and services, including processes to verify that procured items and services meet established requirements, evaluate and select prospective suppliers, and verify that the approved suppliers continue to provide acceptable items and services.

Proposed section IV.A.8, “Criterion 8—Performance: Inspection and Acceptance Testing,” of proposed appendix T to 10 CFR part 50, would include requirements for the applicant to establish and implement processes for performing inspection and acceptance testing for procured items and services.

Proposed section IV.A.9, “Criterion 9—Performance: Maintenance of Structures, Systems, and Components,” of proposed appendix T to 10 CFR part 50, would include requirements for the applicant to establish and implement processes to control the storage of SSCs in accordance with cleanliness and environmental standards. These requirements would ensure that a process is used to prevent foreign material from being introduced to the SSC during storage and to store SSCs in accordance with the required environmental conditions ( e.g., humidity, temperature).

(iii) Assessment

Proposed section IV.A.10, “Criterion 10—Assessment: Management Assessment,” of proposed appendix T to 10 CFR part 50, would include requirements for the applicant to establish and implement processes for management of the organization to assess the continued effectiveness of the QA program.

Proposed section IV.A.11, “Criterion 11—Assessment: Independent Assessment,” of proposed appendix T to 10 CFR part 50, would include requirements for the applicant to establish and implement processes for independent assessment of each aspect of the QA program. These requirements would ensure that those performing these assessments have sufficient authority and freedom from their management and are technically qualified and knowledgeable to perform the assessment.

E. Quality Assurance for Software Used in Design and Analysis, and Digital Items Important to Safety

Proposed section IV.B, “Quality Assurance for Software Used in Design and Analysis, and Digital Items Important to Safety,” of proposed appendix T to 10 CFR part 50, would identify QA criteria for software used for design and analysis of SSCs and for digital items important to safety.

Proposed section IV.B.1 of proposed appendix T to 10 CFR part 50, would require that applicants establish and implement processes within the QA program to ensure that (1) software used in digital items that perform a safety function, (2) software used for design verification for any SSC, and (3) software used for design analysis for any SSC, are documented, managed, and controlled throughout the software life cycle to ensure that the related SSCs perform their intended safety function.

Proposed section IV.B.2 would require the applicant to use appropriate national or internal software engineering standards. Examples of such standards include ASME, Institute for Electrical and Electronics Engineers (IEEE), National Institutes of Standards and Technology (NIST), and American Nuclear Society (ANS).

F. Proposed Conforming Changes to 10 CFR 50.4, 50.34, 50.54, 50.55, 52.79, 53.020, 53.040, 53.460, 53.500, 53.865, 53.1309, 53.1369, 53.1416, and 53.1565

Proposed conforming changes to 10 CFR 50.34(a)(7) would allow a CP applicant who meets the eligibility requirements included in proposed section I of appendix T to 10 CFR part 50, to include in its application for a CP, a QMS that meets proposed appendix T to 10 CFR part 50, as an alternative to satisfying the requirement in 10 CFR 50.34(a)(7) for submittal of a description of the QA program that meets appendix B to 10 CFR part 50.

Proposed conforming changes to 10 CFR 50.34(b)(6)(ii) would allow an OL applicant who meets the eligibility requirements included in proposed section I of appendix T to 10 CFR part 50, to include in its application for an OL, a QMS that meets proposed appendix T to 10 CFR part 50, as an alternative to satisfying the requirement in 10 CFR 50.34(b)(6)(ii) for submittal of a description of the QA program that meets appendix B to 10 CFR part 50.

Proposed conforming changes to 10 CFR 50.34(f)(3)(ii) would add a reference to the proposed appendix T to 10 CFR part 50 for the requirement on ensuring all SSCs important to safety are included in the QA list.

Proposed conforming changes to 10 CFR 50.54(a)(1) would incorporate requirements for:

• Each nuclear power plant or fuel reprocessing plant licensee subject to the QA criteria in proposed appendix T of 10 CFR part 50, to implement, under 10 CFR 50.34(b)(6)(ii) or 52.79, the QMS described or referenced in the safety analysis report, including changes to that report.

• For holders of a COL under 10 CFR part 52, to implement the QMS described or referenced in the safety analysis report applicable to operation 30 days prior to the scheduled date for initial loading of the fuel.

The proposed conforming addition of 10 CFR 50.54(a)(5) would include requirements for changes to a QMS to be submitted to the NRC and receive NRC approval prior to implementation.

Proposed conforming changes to 10 CFR 50.55(f)(1) would incorporate requirements for nuclear power plant or fuel reprocessing plant CP holders subject to the QA criteria in proposed appendix T of 10 CFR part 50, to implement, pursuant to 10 CFR 50.34(a)(7), the QMS described or referenced in the safety analysis report, including changes to that report.

The proposed conforming addition of 10 CFR 50.55(f)(5) would add requirements for changes to a QMS to be submitted to the NRC and receive NRC approval prior to implementation.

The proposed conforming addition of paragraph (b)(7)(iii) to 10 CFR 50.4, “Written communications,” would require a change to the safety analysis report QMS under the proposed 10 CFR 50.54(a)(5) or 10 CFR 50.55(f)(5), or a change to a licensee's NRC-accepted QMS topical report under 10 CFR 50.54(a)(5) or 10 CFR 50.55(f)(5), to be submitted to the NRC's Document Control Desk, with a copy to appropriate Regional Office, and a copy to the appropriate NRC Resident Inspector if one has been assigned to the site of the facility.

Proposed conforming changes to 10 CFR 52.79(a)(25) and (27) would allow a COL applicant who meets the eligibility requirement included in proposed section I of appendix T to 10 CFR part 50, to include in its application for COL, a QMS that meets proposed appendix T to 10 CFR part 50, as an alternative to satisfying the requirement in 10 CFR 52.79(a)(25) and (27) for submittal of a description of the QA program that meets appendix B to 10 CFR part 50.

Proposed conforming changes to 10 CFR 53.020 would modify the definition of QA to align with the definition of QA in the proposed appendix T to 10 CFR part 50.

The proposed conforming addition of paragraph (b)(7)(iii) to 10 CFR 53.040, “Written communications,” would require a change to the safety analysis report QMS under the proposed 10 CFR 53.1565, “Evaluating changes to programs included in licensing-basis information,” or a change to a licensee's NRC-accepted QMS topical report under 10 CFR 53.1565, to be submitted to the NRC's Document Control Desk, with a copy to appropriate Regional Office, and a copy to the appropriate NRC Resident Inspector if one has been assigned to the site of the facility.

Proposed conforming changes to paragraphs (b)(1) and (2) of 10 CFR 53.460, “Safety categorization and special treatments,” would allow, for applicants that meet the eligibility requirements included in proposed section I of appendix T to 10 CFR part 50, the special treatments for safety-related SSCs (under proposed 10 CFR 53.460(b)(1)), and non-safety-related safety-significant SSCs and safety-related SSCs beyond 10 CFR 53.460(b)(1) (under proposed 10 CFR 53.460(b)(2)), to meet applicable QA requirements from proposed appendix T to 10 CFR part 50, as an alternative to these SSCs having to meet the applicable QA requirement in appendix B to 10 CFR part 50.

Proposed conforming changes to paragraph (b) of 10 CFR 53.500, “General siting and siting assessment,” would allow, for applicants that meet the eligibility requirements included in proposed section I of appendix T to 10 CFR part 50, activities performed to identify site characteristics or otherwise needed to determine site-specific contributors to functional design criteria or analysis assumptions under subpart C of 10 CFR part 53 to satisfy the QA requirements from proposed appendix T to 10 CFR part 50, as an alternative for these activities to meet the applicable QA requirement in appendix B to 10 CFR part 50.

Proposed conforming changes to 10 CFR 53.865, “Quality assurance,” for holders of an OL or COL under 10 CFR part 53 that meet the eligibility requirements included in proposed section I of appendix T to 10 CFR part 50, to develop, implement, and maintain a QA program in accordance with proposed appendix T to 10 CFR part 50, as an alternative to appendix B to 10 CFR part 50.

Proposed conforming changes to 10 CFR 53.1309(a)(2)(i) would allow a CP applicant under 10 CFR part 53 who meets the eligibility requirement included in proposed section I of appendix T to 10 CFR part 50, to include in its application for a CP, a QMS that meets proposed appendix T to 10 CFR part 50, as an alternative to satisfying the requirement in 10 CFR 53.109(a)(2)(i) for submittal of a description of the QA program that meets appendix B to 10 CFR part 50.

Proposed conforming changes to 10 CFR 53.1369(l) would allow an OL applicant under 10 CFR part 53 who meets the eligibility requirement included in proposed section I of appendix T to 10 CFR part 50, to include in its application for an OL, a QMS that meets proposed appendix T to 10 CFR part 50, as an alternative to satisfying the requirement in 10 CFR 53.1369(l) for submittal of a description of the QA program that meets appendix B to 10 CFR part 50.

Proposed conforming changes to 10 CFR 53.1416(a)(12) would allow a COL applicant under 10 CFR part 53 who meets the eligibility requirement included in proposed section I of appendix T to 10 CFR part 50, to include in its application for a COL, a QMS that meets proposed appendix T to 10 CFR part 50, as an alternative to satisfying the requirement in 10 CFR 53.1416(a)(12) for submittal of a description of the QA program that meets appendix B to 10 CFR part 50.

Proposed conforming changes to 10 CFR 53.1565(d)(1)(i) would clarify the applicability of QA criteria of appendix B of 10 CFR part 50 for each holder of an OL or COL under 10 CFR part 53, after the Commission makes the finding under 10 CFR 53.1452(g).

Proposed conforming addition of 10 CFR 53.1565(d)(1)(iii) would include requirements for changes to a QMS to be submitted to the NRC and receive NRC approval prior to implementation for each holder of an OL or COL under 10 CFR part 53, after the Commission makes the finding under 10 CFR 53.1452(g).

Proposed conforming changes to10 CFR 53.1565(d)(2) would modify the numbering scheme and clarify the applicability of QA criteria of appendix B to 10 CFR part 50 for each holder of a CP or COL under 10 CFR part 53, before the Commission makes the finding under 10 CFR 53.1452(g).

The proposed conforming addition of 10 CFR 53.1565(d)(2)(ii) would include requirements for changes to a QMS to be submitted to the NRC and receive NRC approval prior to implementation for each holder of a CP or COL under 10 CFR part 53, before the Commission makes the finding under 10 CFR 53.1452(g).

XVII. Background—Updates to Construction Permit Requirements and Related Licenses

The regulations in 10 CFR 50.34 specify the requirements for technical information to accompany an application for a CP or an OL. These regulations were amended in 1968 (33 FR 18610; December 17, 1968) to add paragraph (a) of 10 CFR 50.34 to require an applicant for a CP to submit a preliminary safety analysis report. Paragraph (a) of 10 CFR 50.34 specifies the minimum technical information in the preliminary safety analysis report, including preliminary design information and a description and safety assessment of the site on which the facility is to be located. As stated in the 1968 final rule, the preliminary safety analysis report requirement was “intended to provide early and adequate information which is expected to expedite the processing of CP applications by reducing the time-consuming exchanges between the applicant and the AEC staff required to fill information gaps.” Subsequent changes to 10 CFR 50.34 from 2007 to the present were additions due to new requirements, as well as clarifications and relaxations, but the majority of 10 CFR 50.34 is unchanged since 1968.

The NRC may issue the CP if the agency makes the findings specific to a CP that are listed in paragraph (a) of 10 CFR 50.35, “Issuance of construction permits,” as well as the more general findings for issuance of licenses and permits in 10 CFR 50.40, “Common standards,” and 10 CFR 50.50, “Issuance of licenses and construction permits.” The findings in 10 CFR 50.35(a) stem from the early practices of the AEC, when a “provisional” CP would be issued when an applicant had not submitted all the technical information necessary to complete the application and to approve all proposed design features. Since almost all issued “provisional” CPs were never converted to a “final” CP, the AEC proposed codifying this practice (34 FR 6540; April 16, 1969). The final amendment to the regulations in 10 CFR 50.35 eliminated the term “provisional” CP, but the criteria in 10 CFR 50.35(a) for issuing a CP remained the same as those previously required for a “provisional” CP (35 FR 5317; March 31, 1970). The current regulations for issuing a CP in 10 CFR 50.35(a) have not been modified since 1970.

The NRC issued 10 CFR part 52 on April 18, 1989 (54 FR 15372), to reform the NRC's licensing process for future nuclear power plants. The rule established new approval processes in 10 CFR part 52 for ESPs, standard design certifications, and COLs that were additions to the two-step licensing process that already existed in 10 CFR part 50. This 10 CFR part 52 rule also included processes for standard design approvals and MLs. On August 28, 2007 (72 FR 49352), the NRC issued a final rule with changes to 10 CFR part 52 to clarify the applicability of various requirements to each of the 10 CFR part 52 approval processes.

XVIII. Discussion—Updates to Construction Permit Requirements and Related Licenses

This proposed rule would include updates to the language in 10 CFR 50.34(a) to more clearly link the level of detail required to be submitted with a CP application to the findings the NRC is required to make in 10 CFR 50.35(a), 50.40, and 50.50 to issue a CP. During its review of recent CP applications, the NRC has noted that applicants may provide a higher level of detail for some of the technical areas listed in 10 CFR 50.34(a), and a lower level of detail for others, while still providing sufficient information for the NRC to make the findings required by 10 CFR 50.35(a), 50.40, and 50.50 to issue the CP. The proposed revisions to footnote 1 of 10 CFR 50.34(a) would clarify that the level of detail provided in a preliminary safety analysis report to satisfy the minimum technical requirements in 10 CFR 50.34(a) would be deemed sufficient if the provided information allows the NRC to make the findings required by 10 CFR 50.35(a), 50.40, and 50.50.

This proposed rule would also remove a sentence in 10 CFR 50.34(a)(4) specifying in detail the need to perform loss of coolant accident (LOCA) analyses required in 10 CFR 50.46 and the need for high-point vents in proposed 10 CFR 50.46b, “Acceptance criteria for reactor coolant system venting systems.” The NRC considers that these requirements are already implicitly included in the preceding sentence of 10 CFR 50.34(a)(4), which specifies the need to include an evaluation of “the adequacy of structures, systems, and components provided for the prevention of accidents and the mitigation of the consequences of accidents.” These requirements are also referred to in 10 CFR 50.34(b)(4), and the NRC proposes to remove a sentence from 10 CFR 50.34(b)(4) that similarly specifies in detail the need to perform LOCA analyses required in 10 CFR 50.46.

Conforming changes are also proposed to the similar regulatory text and footnotes in each subpart of 10 CFR part 52 to ensure consistency between the power reactor licensing and approval pathways. These changes would be made in 10 CFR 52.47, which applies to standard design certifications; 10 CFR 52.79, which applies to COLs; 10 CFR 52.137, which applies to standard design approvals; and 10 CFR 52.157, which applies to MLs. This change would reflect a more technology-inclusive approach and remove prescriptive language that could be read to mean that a LOCA analysis methodology is fully developed and validated at the CP stage.

The proposed rule would also adjust the wording in 10 CFR 50.34(a)(1)(ii)(D) to be meaningful for designs with functional containments which are evaluated as a release barrier or series of barriers taken together to perform the containment safety function. Specifically, with the proposed changes, the regulation would no longer prescriptively state that the assumed fission product release be “from the core into the containment” and would be replaced with technology-inclusive language that refers to “leakage rates from potential flow paths” rather than a “containment leak rate.” Light-water reactor designs would continue to use fission product release paths from the core into the containment and containment leak rates. The proposed rule would also revise footnotes 3 and 1 ) as a reference value to remove outdated information regarding recommendations included in a 1959 National Bureau of Standards handbook.

Conforming changes would be made to the similar regulatory text and footnotes in each subpart of 10 CFR part 52 to ensure consistency between the power reactor licensing and approval pathways. Specifically, these proposed changes would be made in 10 CFR 52.17, which applies to ESPs, 10 CFR 52.47, 52.79, 52.137, and 52.157. These changes would reflect a more technology-inclusive approach, eliminate unnecessary exemptions that may otherwise be needed for some designs, and ensure consistency in power reactor applications.

XIX. Background—Alternative Risk-Informed and Performance-Based Acceptance Criteria for 10 CFR Parts 50 and 52

A. Need for Regulatory Flexibility

Many existing NRC regulations include prescriptive acceptance criteria expressed as specific numerical limits ( e.g., temperature, pressure, dose). These criteria were developed based on the state of knowledge and technology at the time the rules were promulgated and do not always reflect the significant advancements in nuclear safety analysis, PRA, and reactor design that have occurred in the decades since. Additionally, the codification of these criteria has limited the ability of licensees and applicants to propose alternative approaches without seeking exemptions, which can introduce cost and regulatory uncertainty.

Licensees and applicants have consistently identified unduly prescriptive requirements as a deterrent to innovation. This has been particularly challenging for U.S. companies developing new reactor designs and seeking to compete in global markets. Nonetheless, the NRC has determined that a top-down approach—modifying individual prescriptive requirements throughout 10 CFR parts 50 and 52—would be resource-intensive and could have unintended consequences, especially for the licensing bases of currently operating reactors. Therefore, instead of a top-down approach, the NRC is proposing to expand the use of risk-informed and performance-based alternatives to existing prescriptive requirements.

B. Enabling Risk-Informed and Performance-Based Alternatives

The NRC has long supported the use of risk-informed and performance-based approaches in its regulatory decision-making, as reflected in the Commission's policy statements and strategic goals, particularly, SRM-SECY-98-144, “Staff Requirements—SECY-98-144—White Paper on Risk-Informed and Performance-Based Regulation,” dated March 1, 1999. Since then, the agency has encouraged the use of such approaches to improve regulatory decision-making, enhance safety, and reduce unnecessary regulatory burden.

The proposed rule would build on this foundation by providing a structured pathway for licensees and applicants to propose alternative acceptance criteria that would be tailored to demonstrate the safety of their specific technologies without the need for exemptions. Additionally, the proposed rule would further utilize risk-informed and performance-based methodologies to update appendix A to 10 CFR part 50 to clarify the application of general design criteria (GDCs) during the licensing of new LWR designs.

XX. Discussion—Alternative Risk-Informed and Performance-Based Acceptance Criteria for 10 CFR Parts 50 and 52

The NRC is proposing to add new, standalone regulations, 10 CFR 50.220 and 10 CFR 52.220, entitled “Use of risk-informed and performance-based alternatives to acceptance criteria,” to allow licensees and applicants to voluntarily submit and use technology-inclusive, risk-informed, or performance-based acceptance criteria as alternatives to existing prescriptive requirements. These new provisions would support the expanded and accelerated use of acceptance criteria reflective of innovative nuclear technologies without the need for exemptions, while continuing to ensure reasonable assurance of adequate protection of public health and safety.

In addition, the NRC is proposing to update appendix A to 10 CFR part 50 to clarify that (1) deviations from GDCs could be identified and justified within licensing submittals, with no need for a separate exemption request; and (2) demonstrating compliance with Criterion 28, “Reactivity limits” (GDC 28), of appendix A to 10 CFR part 50 could be based on a different design basis accident than the control rod ejection or control rod drop accident.

The proposed 10 CFR 50.220 and 10 CFR 52.220 would address regulatory inefficiencies and foster innovation by offering a flexible, voluntary alternative to the current approach, which in many instances relies on prescriptive requirements that applicants must seek exemptions from when proposing to adopt innovative methodologies. The NRC has identified several cases in which this type of framework could have enabled more timely and efficient regulatory decisions. These experiences highlight the value of reducing the number of exemptions and rulemakings required, thereby improving efficiency and supporting the deployment of new technologies. The proposed rule would also provide opportunities for increased operational flexibility at existing facilities that choose to propose and adopt alternative criteria.

The proposed approach would enhance regulatory flexibility, efficiency, and reliability by providing a voluntary pathway for the use of alternative criteria. This approach would maintain the existing licensing basis for currently operating reactors and potential restart units, thereby avoiding unintended impacts associated with a broad, top-down revision of regulatory requirements. Additionally, the proposed rule would establish a more transparent and structured process for NRC review and acceptance of alternative criteria.

This approach would be consistent with the NRC's commitment to enabling the safe use of nuclear technology for the benefit of society, while maintaining reasonable assurance of adequate protection of public health and safety. It would also be consistent with the approach of previous NRC rulemakings that provided voluntary pathways for risk-informed and performance-based alternatives to existing requirements, such as the promulgation of 10 CFR 50.69, “Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors” (69 FR 68008; November 22, 2004) and 10 CFR part 53 (91 FR 15696; March 30, 2026).

The NRC recognizes that successful implementation of this proposed approach would require broad and flexible guidance to accommodate the range of potential alternative criteria. As a result, NRC encourages the increased

Looking forward, the NRC would maintain a record of NRC-approved alternative acceptance criteria, along with references to the associated bases for approval to support streamlined use of approved alternative acceptance criteria by potential applicants and serve as a means of regulatory recordkeeping for the agency.

The proposed changes would include revising appendix A to 10 CFR part 50 to clarify that exemptions are not required for deviations from the GDCs. Instead, such deviations could be identified and justified directly within the licensing application. While this clarification would not change the NRC's review of the justification itself, it would reduce regulatory burden for applicants proposing innovative designs in which certain GDCs may be tailored to be better risk-informed.

In addition, the NRC is proposing to revise GDC 28 of appendix A to 10 CFR part 50 to allow applicants to propose and justify alternative design basis accidents for reactivity control, rather than prescriptively requiring evaluation of control rod ejection or control rod drop accidents. Although applicants and licensees may be able to accomplish this through the flexibilities in the proposed 10 CFR 50.220 and 52.220, this change would more explicitly enable the use of design-specific accident scenarios and support removal of unnecessary conservatism in safety analyses without the need to provide the information that would be required in proposed 10 CFR 50.220(b) and 52.220. The NRC is issuing, for public comment along with this proposed rule, DG-1464, “Guidance for Content of Applications Under 10 CFR 50.220 and 52.220 Proposing Risk-Informed and Performance-Based Alternative Acceptance Criteria,” which would be used to support applicants in determining appropriate design basis accidents based on credible, realistic risks.

XXI. Background—Establishing Thresholds for Changes to Reactor Designs During Construction and Operation Under 10 CFR Parts 52 and 53

Section 5(f) of E.O. 14300 directs the NRC to establish stringent thresholds for circumstances in which the NRC may demand changes to a reactor design once construction of the reactor is underway. In response, the NRC is proposing to raise the threshold for changes required during construction by eliminating “increased standardization” as a criterion for Commission-directed modification of design certification information on either a plant-specific or generic basis. In addition, this proposed rule would provide additional flexibility and reduce unnecessary regulatory burden in the regulations governing licensee-requested changes during construction under 10 CFR parts 52 and 53. The NRC is also proposing changes to 10 CFR parts 52 and 53 to provide flexibility and efficiencies for licensee-requested changes during operation.

A. Development of Tiers of Information and Processes for Changes and Departures in Design Certification Rules

In 1987, the NRC issued a policy statement on nuclear power plant standardization (52 FR 34884; September 15, 1987). In 1989, 10 CFR part 52 was issued (54 FR 15372; April 18, 1989), which established that design certification would be accomplished by rulemaking. The NRC ultimately adopted a two-tiered system for design information. These tiers are designated in a design control document (DCD), which the NRC incorporates by reference into its regulations, and each information tier is subject to a specified process for changes and departures from design certification information.

Tier 1 information is the portion of the DCD that is approved and certified. It includes definitions and general provisions; design descriptions; inspections, tests, analyses, and acceptance criteria (ITAAC); significant site parameters; and significant interface requirements. Tier 1 design descriptions were intended to be applicable for the life of the facility. Tier 2 and Tier 2* information is the portion of the DCD that is approved but not certified. Tier 2 includes information like that found in a final safety analysis report (FSAR) for 10 CFR part 50 licenses. Tier 2* was created to minimize information in Tier 1 while requiring that this information could not be changed without prior NRC approval. If the Tier 2* designation were not available, this information would have been designated Tier 1. While the Tier 2* category was used in the first five design certifications, from the U.S. Advanced Boiling Water Reactor (ABWR) to the Economic Simplified Boiling-Water Reactor (ESBWR), the later APR1400 and NuScale DCDs did not designate any information as Tier 2*.

Each design certification in appendix A, “Design Certification Rule for the U.S. Advanced Boiling Water Reactor,” appendix D, “Design Certification Rule for the AP1000 Design,” appendix E, “Design Certification Rule for the ESBWR Design,” appendix F, “Design Certification Rule for the APR1400 Design,” and appendix G, “Design Certification Rule for NuScale,” of 10 CFR part 52 includes a section VIII, “Processes for Changes and Departures,” that specifies processes to change Tier 1, Tier 2, and, where applicable, Tier 2* information. The requirements in section VIII of these appendices are essentially identical except for certain certified designs that do not have information designated as Tier 2*. Under these section VIII requirements, plant-specific changes or departures from Tier 1 require an exemption. Plant-specific departures from Tier 2* require NRC approval by license amendment. Changes or departures from Tier 2 information are evaluated using a process like the one provided for 10 CFR part 50 licensees in 10 CFR 50.59, which provides criteria for determining whether a licensee-initiated change requires prior NRC approval. This similar departure process for Tier 2 information is, therefore, often described as a 10 CFR 50.59-like process for COL holders referencing a design certification. The change processes in section VIII sought to balance standardization with flexibility and apply during both construction and operation.

For currently certified designs, Tier 2* is defined in section II.F of appendices A, D, and E to 10 CFR part 52. Appendices F and G to 10 CFR part 52 do not contain Tier 2* information. Tier 2* is the portion of Tier 2 information designated with brackets, italicized text, and an asterisk in the generic DCD. Per sections VIII.B.6.b and VIII.B.6.c of appendices A, D, and E to 10 CFR part 52, any licensee who references these appendices may not depart from Tier 2* matters without prior NRC approval and any request for such a departure will be treated as a request for a license amendment under 10 CFR 50.90. Sections VIII.B.6.b and VIII.B.6.c of appendices A, D, and E to 10 CFR part 52 list all of the Tier 2* matters for the respective designs. The Tier 2* matters listed in section

Another key aspect of the standard design certification is finality. Issuance of the design certification rule allows the design to be incorporated by reference into a COL application. Issue finality rules limit the types of changes that may be imposed on the certification information. During the COL application review, design information codified by rule is not subject to NRC review, but the NRC staff would review applicant-requested departures from the certified design that require NRC approval. The scope of a hearing for a COL application that references a certified design does not include the design certified by NRC rule but would encompass any departures from the certified design requiring NRC approval.

The final rule promulgating 10 CFR part 53 established a risk-informed, performance-based, and technology-inclusive regulatory framework for commercial nuclear plants, including advanced reactor designs. Part 53 of 10 CFR dispensed with the Tier 1 and Tier 2 terminology. Rather, 10 CFR 53.1525, “Revising certification information within a design certification rule,” uses the term “certification information” in place of Tier 1. Information that is “not certification information” is equivalent to Tier 2 information under 10 CFR part 52. The change control processes for “certification information” and “not certification information” in 10 CFR part 53 are similar to those for Tier 1 and Tier 2 in 10 CFR part 52.

B. Licensing Experience and Improvement Initiatives Regarding Information Designation and Change Processes for Design Certifications

The NRC has periodically considered improvements to the effectiveness and efficiency of information designations for standard design certifications and associated change processes. Many of these improvements, developed through internal and external reviews and interactions, are addressed in this proposed rule.

The NRC staff internally reviewed and considered improvements to the content of Tier 2* and Tier 1 i

This text is long and has been trimmed here. Open the source document for the complete record.