SUPPLEMENTARY INFORMATION:

Executive Summary

A. Need for the Regulatory Action

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

B. Major Provisions

Major provisions of this proposed rule, supported by accompanying draft guidance (if applicable), include changes in the following areas:

a. Part 21 Clarifications (10 CFR Parts 21, 50, and 53)

b. Reporting Requirements for Nonemergency Events for Nuclear Power Plants (10 CFR Part 50)

c. Decommissioning Licensing Enhancements (10 CFR Parts 20, 50, 52, and 53)

d. Financial Qualifications and Decommissioning Trust Fund Use (10 CFR Parts 50, 52, and 53)

e. Risk Informed Seismic Design (10 CFR Part 50)

f. Probabilistic Risk Assessment Requirements (10 CFR Part 50)

g. Backfitting and Issue Finality Provisions (10 CFR Parts 50, 52, 53, 70, and 72)

h. Notification of Initial Fuel Load (10 CFR Parts 52 and 53)

i. Requirements for Evaluation of Generic Issues and Operating Experience (10 CFR Part 52)

j. Standard Design Approvals (10 CFR Parts 50, 52, and 53)

k. Environmental Reviews of Early Site Permits (10 CFR Parts 52 and 53)

l. Post-TMI Requirements (10 CFR Part 50)

m. Updates to the Operator Licensing Programs (10 CFR Parts 50, 53, and 55)

n. International Atomic Energy Agency (IAEA) Notifications (10 CFR Part 75)

o. Record Keeping and Reporting Requirements (10 CFR Parts 50, 52, 53, 70, and 72)

C. Costs and Benefits

The NRC prepared a draft regulatory analysis to determine the expected quantitative costs and benefits of this proposed rule and associated draft guidance as well as qualitative factors to be considered in the NRC's rulemaking decision. The conclusion from the analysis is that this proposed rule and associated draft guidance would result in net averted costs ( i.e., savings) to the industry and the NRC, over the next 30 years, ranging from $311 million using a 7-percent discount rate to $411 million using a 3-percent discount rate. For the industry, the net averted cost is estimated at $305 million (7 percent net present value (NPV) and $398 million (3 percent NPV). For the NRC, the net averted cost is estimated at $5.50 million (7 percent NPV) and $13.9 million (3 percent NPV). On an annualized basis, the net averted cost to the industry and the NRC would be about $22.2 million per year at a 7 percent discount rate and $15.1 million per year at a 3 percent discount rate.

The draft regulatory analysis also considers qualitative factors, such as regulatory efficiency. These benefits would result from clarifications, administrative changes, and streamlining of processes (such as notifications).

For more information, please see the draft regulatory analysis (available in the NRC's Agencywide Documents Access and Management System (ADAMS) Accession No. ML26191A071).

Table of Contents

I. Obtaining Information and Submitting Comments

II. Executive Order 14300: Ordering the Reform of the Nuclear Regulatory Commission

III. Background

IV. Discussion

V. Specific Requests for Comments

VI. Regulatory Flexibility Certification

VII. Regulatory Analysis

VIII. Backfitting and Issue Finality

IX. Cumulative Effects of Regulation

X. Plain Writing

XI. National Environmental Policy Act

XII. Paperwork Reduction Act

XIII. Executive Orders

A. Executive Order 12866: Regulatory Planning and Review (as Amended by Executive Order 14215, Ensuring Accountability for All Agencies)

B. Executive Order 14154: Unleashing American Energy

C. Executive Order 14192: Unleashing Prosperity Through Deregulation

D. Executive Order 14267: Reducing Anti-Competitive Regulatory Barriers

E. Executive Order 14270: Zero-Based Regulatory Budgeting To Unleash American Energy

XIV. Compatibility of Agreement State Regulations

XV. Availability of Guidance

XVI. Availability of Documents

( printed page 60703)

I. Obtaining Information and Submitting Comments

A. Obtaining Information

Please refer to Docket ID NRC-2025-1138 when contacting the NRC about the availability of information for this action. You may obtain publicly available information related to this action by any of the following methods:

• Federal Rulemaking Website: Go to www.regulations.gov and search for Docket ID NRC-2025-1138.
• NRC's Agencywide Documents Access and Management System (ADAMS): You may obtain publicly available documents online in the ADAMS Public Documents collection at 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.
• 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.
• Public Meeting: The NRC may conduct a public meeting to describe the proposed amendments and answer questions from the public on this proposed rule. If the NRC determines it will hold a public meeting, NRC will publish a notice of the location, time, and agenda of the meeting on the NRC's public meeting website within 10 calendar days of the meeting. Stakeholders should monitor the NRC's public meeting website for information about the public meeting at: www.nrc.gov/​public-involve/​public-meetings/​index.cfm.

B. Submitting Comments

Comments must be submitted using www.regulations.gov by 11:59 p.m. eastern time on November 9, 2026. Please include Docket ID NRC-2025-1138 in your comment submission.

The NRC cautions you not to include identifying or contact information that you do not want to be publicly disclosed in your comment submission. The NRC will post all comment submissions at 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 E.O. 14300, “Ordering the Reform of the Nuclear Regulatory Commission.” Section 5, “Reforming and Modernizing the NRC's Regulations,” of that E.O. requires the NRC to undertake a review and wholesale revision of its regulations and guidance documents as guided by the policies set forth in section 2 of the E.O. This rulemaking addresses the required review and wholesale revision of the NRC's regulations and guidance documents specified in the main body of section 5 of E.O. 14300.

III. Background

Over the decades, under its authority to make, promulgate, issue, rescind, and amend such rules and regulations as may be necessary, the NRC has developed a comprehensive and evolving regulatory framework to ensure that licensee programs at nuclear facilities continue to provide reasonable assurance of adequate protection to the health and safety of the public and common defense and security. This proposed rule seeks to modernize the NRC's regulations across 10 CFR parts 20, 21, 50, 52, 53, 55, 70, 72, and 75 with a focus on reducing or eliminating duplicative or unnecessary requirements, providing flexibility, and implementing risk-informed, performance-based alternatives where appropriate with respect to reactor licensing, decommissioning, and operational oversight.

The NRC recently issued a final rule establishing 10 CFR part 53, “Risk-Informed, Technology-Inclusive Regulatory Framework for Commercial Nuclear Power Plants” (91 FR 15696, March 30, 2026). Part 53 sets forth a technology-inclusive regulatory framework as an alternative to the NRC's current application and licensing requirements, which were primarily developed to address license requests concerning light water-cooled reactors and operational requirements for those types of reactors. As discussed in Section IV of this document, proposed changes would apply to licensees and applicants under 10 CFR parts 50, 52, and 53, as applicable. The proposed changes would provide enhanced regulatory flexibility for both current and future licensees, streamline administrative processes, and ensure that NRC requirements remain effective, efficient, and responsive to credible risks.

The NRC prepared an unofficial redline strikeout version of the proposed changes to regulatory text that is intended to help the reader identify the proposed changes. The unofficial redline strikeout version of this proposed rule is publicly available and is listed in the “Availability of Documents” section.

For the convenience of the reader, this proposed rule has been organized as follows within each section:

a. 10 CFR Parts 21, 50, and 53—Part 21 Clarifications

b. 10 CFR Part 50—Reporting Requirements for Nonemergency Events for Nuclear Power Plants

c. 10 CFR Parts 20, 50, 52, and 53—Decommissioning Licensing Enhancements

d. 10 CFR Parts 50, 52, and 53—Financial Qualifications and Decommissioning Trust Fund Use

e. 10 CFR Part 50—Risk Informed Seismic Design

f. 10 CFR Part 50—Probabilistic Risk Assessment Requirements

g. 10 CFR Parts 50, 52, 53, 70, and 72—Backfitting and Issue Finality Provisions

h. 10 CFR Parts 52 and 53—Notification of Initial Fuel Load

i. 10 CFR Part 52—Requirements for Evaluation of Generic Issues and Operating Experience

j. 10 CFR Parts 50, 52, and 53—Standard Design Approvals

k. 10 CFR Parts 52 and 53—Environmental Reviews of Early Site Permits

l. 10 CFR Part 50—Post-TMI Requirements

m. 10 CFR Parts 50, 53, and 55—Updates to the Operator Licensing Programs

n. 10 CFR Part 75—IAEA Notifications

o. 10 CFR Parts 50, 52, 53, 70, and 72—Record Keeping and Reporting Requirements

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

The NRC is proposing revisions to 10 CFR part 21, “Reporting of Defects and Noncompliance,” for reporting safety defects. The regulations in 10 CFR part 21 establish procedures and requirements for the implementation of section 206 of the Energy Reorganization Act of 1974 (ERA). At a ( printed page 60704) high-level, 10 CFR part 21 addresses risks from basic components that have been identified to contain a defect, which could result in a plant failing to meet its licensing basis. The requirements and procedures of 10 CFR part 21 contribute to public health and safety by ensuring the Commission is adequately informed of any loss of safety function to the extent that there is a major reduction in the degree of protection provided to the public health and safety. Under 10 CFR part 21, directors and responsible officers of firms and organizations building, operating, or owning NRC-licensed facilities are required to report failures to comply with regulatory requirements relating to substantial safety hazards and defects in components that may result in a substantial safety hazard. This regulation also applies to directors and responsible officers of firms and organizations supplying safety-related components, as well as those providing safety-related design, testing, inspection, and consulting services. Part 21 also requires these entities to adopt procedures to ensure that safety-related defects and noncompliance are brought to the attention of responsible officers and directors (or their designees). In turn, the responsible officers and directors (or their designees) are required to notify the Commission by filing an initial report followed by a written report regarding the defect or noncompliance.

For facilities licensed under 10 CFR parts 40, 70, and 21 has been applicable since its publication in the final rule “Reports to the Commission Concerning Defects and Noncompliance” (42 FR 28891; June 6, 1977). Since then, the lack of clarity in 10 CFR part 21 has presented challenges to licensees, vendors, and the NRC staff. The rule's preamble discusses the broad, overarching nature of the regulations in 10 CFR part 21. In particular, it identifies the need for additional information to be accumulated after the rule has been implemented, with a recommendation that additional guidance clarify the regulation once experience has been gained. In response, the NRC published NUREG-0302, “Remarks Presented (Questions/Answers Discussed) at Public Regional Meetings to Discuss Regulations (10 CFR part 21) for Reporting of Defects and Noncompliance,” to provide further clarity, including thresholds for the substantial safety hazard criterion of “moderate exposure to, or release of, licensed material.” The information presented in NUREG-0302 does not provide specific guidance to applicants or licensees on how to comply with 10 CFR part 21; rather, it is a collection of questions and answers from a series of public meetings regarding 10 CFR part 21 requirements held from July 12-26, 1977. Additionally, the NRC staff has not provided detailed guidance to clarify the nexus between 10 CFR parts 21 and 70 requirements including the updated regulatory requirements in subpart H to 10 CFR part 70, “Additional Requirements for Certain Licensees Authorized to Possess a Critical Mass of Special Nuclear Material” issued in September 2000, such as which types of structures, systems or components at a facility licensed under 10 CFR part 70 would be subject to the basic component requirements in 10 CFR part 21.

The NRC first amended 10 CFR part 21 in 1978 in order to exempt commercial-grade items (CGIs) from 10 CFR part 21 regulations until the items were dedicated for use as basic components (43 FR 48621; October 19, 1978). This rulemaking established the definition of CGIs and the commercial-grade dedication process, which would still be applicable to reactor and nonreactor facilities. In 1991, the NRC issued Information Notice 91-39, “Compliance with 10 CFR part 21, `Reporting of Defects and Noncompliance,' ” to inform NRC materials licensees of their responsibilities under 10 CFR part 21. Additionally in 1991, 10 CFR part 21 was amended to clarify the definitions of “defect” and “basic component” as well as to add the requirement for the retention of 10 CFR part 21 records (56 FR 36081; July 31, 1991). This rulemaking did not clarify the definition for, nor did it give specificity for the identification of, nonreactor basic components. In 1995, the NRC updated 10 CFR part 21 again (60 FR 48369; September 19, 1995) to provide flexibility in procuring CGIs for safety-related equipment and services to nuclear power plants; however, the issue of CGIs with respect to nonpower reactor licensees was not changed.

In 2010 and 2011, the NRC's Office of the Inspector General (OIG) performed two audits related to 10 CFR part 21. These audits resulted in 15 recommendations, most of which related to clarifying 10 CFR part 21.^[1] One of the issues identified was that there was a lack of clarity in the definition of “basic component” for nonreactor facilities and activities that made it difficult to interpret the applicability of the rule to 10 CFR part 70 licensees. This lack of clarity was illustrated by the submittal of alternative definitions by applicants and licensees to clarify the term “basic component.” The NRC staff engaged stakeholders in multiple public meetings and also undertook a series of site visits at fuel cycle facilities. These interactions uncovered that there were widely differing interpretations of the meaning of “basic component” at different fuel cycle facilities. Specifically, at some facilities the term was interpreted to include all systems, structures, and components designated as item relied on for safety (IROFS) in accordance with § 70.61(e), whereas at other facilities the term was interpreted as applying only to those IROFS that are the sole item relied on to prevent high consequence events as defined in § 70.61(b) or to mitigate their effects.

In 2015, the NRC staff issued proposed changes to 10 CFR part 21 and an assessment of the areas that need clarification in “Final Regulatory Basis to Clarify 10 CFR part 21, `Reporting of Defects and Noncompliance' ” (ML15152A457). This final regulatory basis proposed clarifying the 10 CFR part 21 regulations through both rulemaking and the development of regulatory guidance. However, these rulemaking actions were subsequently discontinued (ML16028A189); therefore, changes that would have clarified application of 10 CFR part 21 at fuel facilities were not implemented.

As part of efforts to address E.O. 14300, the NRC has reassessed whether certain potential changes identified in these prior activities would be appropriate. This rulemaking proposes the following revisions to 10 CFR part 21, paragraph (e) of § 50.55, “Conditions of construction permits, early site permits, combined licenses, and manufacturing licenses,” and § 53.605, “Reporting of defects and noncompliance,” as a result of this reassessment and broader assessment conducted pursuant to E.O. 14300. The proposed changes would include consolidating specific requirements in §§ 50.55(e) and 53.605 into 10 CFR part 21, allowing deletion of the redundant regulations in §§ 50.55(e) and 53.605.

The NRC proposes to amend five definitions: “dedicating entity,” “basic component,” “dedication,” “commercial grade item,” and “notification.” One proposed change would revise the definition of “dedicating entity” to remove regulatory constraints that could limit competition in accordance with E.O. 14267, ( printed page 60705) “Reducing Anti-Competitive Regulatory Barriers.” This revision would delete the description of who can perform the dedication; specifically, it would remove references to a third-party dedicating entity. This revision would also separate the definition of dedicating entity based on the type of regulated facilities and the required quality assurance program. Specifically, the definition has been separated into: (1) facilities licensed under 10 CFR parts 50, 52, and 53; (2) 10 CFR part 70 licensees and other licensees, except for plutonium processing and fuel fabrication plants or spent fuel reprocessing facilities, that are required to develop an integrated safety analysis in accordance with 10 CFR part 70, subpart H; and (3) plutonium processing and fuel fabrication plants or spent fuel reprocessing facility licensees licensed pursuant to 10 CFR part 70 that are required to develop an integrated safety analysis in accordance with subpart H to 10 CFR part 70. For the proposed definition of “dedicating entity” for plutonium processing and fuel fabrication plants or spent fuel reprocessing facilities, the definition would be revised to conform with a proposed change to § 70.22(f) ^[2] to allow organizations with a dedication process compliant with a 10 CFR part 50, appendix B or a proposed appendix T, “Streamlined Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Facilities,” ^[3] quality assurance program to perform dedication activities. In addition, the definition of “dedicating entity” would be modified to remove the reference to § 21.21(c) in order to not duplicate the requirements of § 21.21(c). Another proposed revision would modify the definition of “basic component” to (1) provide clarity to fuel facility licensees that are required to develop an integrated safety analysis to focus on equipment that is a sole engineered IROFS and is essential to preventing or mitigating an intermediate-consequence event and a high-consequence event including an accidental criticality; (2) provide clarity to plutonium processing and fuel fabrication plants or spent fuel reprocessing facilities that are required to develop an integrated safety analysis to focus on equipment that is relied on to prevent or mitigate high-consequences to individuals located beyond the controlled area boundary from a release of fission products and transuranics; (3) provide consistency with the definition of safety-related SSCs under 10 CFR part 53; and (4) allow items designed and manufactured under a quality assurance program that is complying with the proposed appendix T, to 10 CFR part 50 to be included in the definition of a basic component. The proposed changes to the “dedication” and “commercial grade item” definitions would (1) clarify that nuclear material users are exempt from the commercial grade dedication process; (2) add a reference to 10 CFR part 52 to address concerns raised after the 2007 10 CFR part 52 rulemaking, in which the NRC recognized that conforming changes to definitions in 10 CFR part 21 were not included to address 10 CFR part 52; (3) clarify that items that have successfully undergone dedication are deemed equivalent to items designed and manufactured under a quality assurance program that is complying with the proposed appendix T; and (4) add a reference to 10 CFR part 53. The proposed change to “notification” describes the primary method that the communication to the NRC should occur.

This proposed rule would remove 10 CFR part 76, “Certification of Gaseous Diffusion Plants,” from the scope of 10 CFR part 21. This proposed rule would also remove the “Initial notification by facsimile” requirement in § 21.21(d)(3)(i) and provide an updated email submittal option. Overall, these proposed revisions would increase clarity and reduce the regulatory burden.

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

The NRC is proposing to amend its regulations related to reporting nonemergency events at nuclear power plants. The NRC would retain reporting criteria that are important to plant safety and maintaining appropriate NRC awareness of nonemergency events at nuclear power plants. The NRC would eliminate criteria that are no longer needed. This proposed rule would also allow for alternative reporting methods for certain nonemergency events. The following subsections provide further background on the specific reporting requirements that would be revised or eliminated.

A. Immediate Notification Requirements Under § 50.72

In a 1980 final rule, “Immediate Reporting of Significant Events at Operating Nuclear Power Reactors” (45 FR 13434; February 29, 1980), the NRC established in § 50.72, “Immediate notification requirement for operating nuclear power reactors,” immediate reporting requirements for events considered significant. In the preamble to that rule, the Commission stated that the capability of the NRC to make timely decisions and to provide adequate assurances regarding actual or potential threats to public health and safety, depends heavily on the rapidity with which significant events are communicated by nuclear power reactor licensees to the NRC. The majority of events occurring throughout the nuclear industry pose little or no serious or immediate threats to the public health and safety; however, certain events do pose such threats or generate fear or unusual concern. The Commission also stated that the NRC has an important obligation to collect facts quickly and accurately about significant events, assess the facts, take necessary action, and inform the public about the extent of the threat, if any, to public health and safety. Not only must the NRC act promptly to prevent or minimize possible injury to the public, but it must also take appropriate action to alleviate fear or concern created as a result of such events.

Accordingly, the NRC required licensees to report 12 types of significant events to the NRC within 1 hour. These included serious events that could result in an impact on public health and safety, such as a nuclear power plant being in an uncontrolled condition, the exceedance of a safety limit, or an uncontrolled release of radioactivity.

Experience with the reporting requirements in § 50.72 led to the issuance of proposed amendments on December 21, 1981, “Immediate Notification Requirement for Operating Nuclear Reactors” (46 FR 61894). Public comments on the proposed rule suggested that the regulations should be revised to clarify reporting criteria and to require early reports only on those matters of value to the exercise of the NRC's responsibilities. The Commission amended the reporting requirements of § 50.72 in the 1983 final rule, “Immediate Notification Requirements of Significant Events at Operating Nuclear Power Reactors” (48 FR 39039; ( printed page 60706) August 29, 1983). The amendments clarified the list of reportable events with the goal of providing the Commission with more useful reports about the safety of operating nuclear power plants. The Commission also created a category of “nonemergency” event reports.

Subsequently, the Commission amended the reporting requirements in § 50.72 in a final rule published on September 10, 1992 (57 FR 41378). The Commission determined that some of the event reports were not necessary for the NRC to perform its safety mission and that continued reporting of those events would not contribute useful information to the operating reactor events database. The Commission also considered that unnecessary reports would continue to consume licensee and NRC resources that could be better applied elsewhere. As a result, the NRC removed reports for events primarily involving invalid engineered safety feature actuations that it determined were of little or no safety significance.

In 2000, the Commission significantly modified the reporting requirements of § 50.72 in a final rule, “Reporting Requirements for Nuclear Power Reactors and Independent Spent Fuel Storage Installations at Power Reactor Sites” (65 FR 63769; October 25, 2000). These amendments divided the nonemergency event reporting requirements into 1-hour, 4-hour, and 8-hour notifications instead of only 1-hour and 4-hour reporting requirements. The 1-hour reports were intended to “capture those events where there may be a need for immediate action by the NRC to protect public health and safety.” For 4-hour reports, the information was determined to be necessary in a prompt manner because the NRC may need to respond to heightened public concern or take a reasonably prompt action, such as partial activation of the agency's response plan. The 8-hour reports were intended to capture those events for which the NRC may need to take action within about a day, such as initiation of a special inspection.

B. Recent Rulemaking Activity

On August 2, 2018, Bill Pitesa, on behalf of the Nuclear Energy Institute (NEI), submitted a petition for rulemaking (PRM) under § 2.802, “Petition for rulemaking—requirements for filing,” requesting that the NRC amend § 50.72 to remove all nonemergency event notification requirements. In accordance with § 2.803, “Petition for rulemaking—NRC action,” the NRC docketed this petition for rulemaking as PRM-50-116 (the petition) and reviewed the petition under § 2.803(h)(1). The NRC staff provided a rulemaking plan to the Commission under SECY-20-0109, “Petition for Rulemaking and Rulemaking Plan on Immediate Notification Requirements for Nonemergency Events (PRM-50-116; NRC-2018-0201),” dated November 30, 2020, that recommended evaluating the nonemergency event reporting requirements of § 50.72(b), “Non-emergency events.” In the staff requirements memorandum (SRM) SRM-SECY-20-0109, “Petition for Rulemaking and Rulemaking Plan on Immediate Notification Requirements for Nonemergency Events,” dated July 28, 2021, the Commission directed the NRC staff to proceed with a rulemaking on the immediate notification requirements for nonemergency events in § 50.72(b) and approved the closure of the docket for PRM-50-116. The Commission provided additional direction to the staff in SRM-SECY-20-0109 that (1) changes that shift the responsibility for gathering, verifying, and communicating information on nonemergency events from licensees to resident inspectors should not be pursued, and (2) the staff should discontinue this effort if it determines at any point that the costs are not justified by the benefits or that the rulemaking would unacceptably degrade the NRC's situational awareness.

Subsequently, the NRC published in the Federal Register a regulatory basis to support rulemaking (87 FR 67571; November 9, 2022) with a comment period ending on January 9, 2023. In the regulatory basis, the NRC recommended maintaining five nonemergency event reporting requirements, eliminating six nonemergency event reporting requirements, revising guidance for two nonemergency event reporting requirements, and adding an alternative reporting method. The NRC held public meetings on November 4, 2021, on December 9, 2021, and on December 13, 2022. The meeting summaries are available as indicated in the “Availability of Documents” section of this document. The NRC extended the public comment period for the regulatory basis to January 31, 2023 (87 FR 79258; December 27, 2022). The NRC received twelve comment submissions on the regulatory basis, all of which were considered in the development of this proposed rule and are available in regulations.gov under Docket ID NRC-2020-0036.

C. NRC Observations on Stakeholder Feedback on the Regulatory Basis

The NRC received stakeholder feedback in several technical areas on the regulatory basis. The public comment submissions are listed in the following table, including associated ADAMS Accession Numbers. The NRC compiled the annotated comment submissions into a single document, which may be accessed within ADAMS using accession number ML23318A475. The NRC reviewed these submissions, several of which resulted in additional analysis and informed the development of this proposed rule. The additional analysis did not result in changes to the recommendations from the regulatory basis.

The NRC parsed each submission, depending on its length and complexity, into one or more individual comments reflecting discrete statements. The following discussion presents the NRC's summaries of and responses to these comments.

Summary of Comments and NRC Responses

Comment 1 (001-1, 003-1, 004-1, 005-1, 006-1, 007-1, 011-1): The NRC received several comments in support of the proposed rulemaking and encouraging the NRC to consider further changes to current event notification requirements.

NRC Response: The NRC agrees, in part, with the comments. The NRC is proposing changes consistent with the analysis the regulatory basis. The NRC may consider additional changes to event notification requirements in the future.

Accordingly, the Commission did not change the NRC staff's recommendations in the regulatory basis in response to these comments.

Comment 2 (002-01, 002-02, 002-03, 008-1, 009-7, 009-8, 009-9, 012-2, 012-3): Several comments reiterated the usefulness of event notifications for the purposes of awareness and the ability for stakeholders to consider taking actions independent of any NRC actions. The proposed elimination of prompt nonemergency event notifications would prevent the public from being adequately informed about many issues at the plant until 60 days following an event, and the methodology in the regulatory basis failed to analyze the importance of prompt communications for effective responses to public inquiries. Right now, the reports are made publicly available on the NRC website, which is the timeliest source of information provided to the public. The State, local, and Tribal governments are the most impacted, as members of the public are more likely to reach out to trusted State and local officials to understand the potential impacts of events. Absent receiving these prompt public notifications, State and local officials must expend valuable time and resources to reach out to licensees and the NRC, thereby delaying response to the public.

NRC Response: The NRC agrees, in part, with the comments. Event notifications may be useful for awareness of external parties; however, the situational awareness of State and local officials for nonemergency events is outside the scope of the NRC's regulatory authority. The NRC is proposing to no longer compel event notification reports that are not important to plant safety and are unnecessary to maintain appropriate NRC awareness of nonemergency events at nuclear power plants. The comments did not provide any examples of NRC situational awareness that would be degraded in the absence of the nonemergency event notifications recommended for elimination through the rulemaking.

Accordingly, the Commission did not change the bulk of the NRC staff's recommendations in the regulatory basis in response to these comments. The Commission did reconsider removal of reporting criterion § 50.72(b)(3)(xiii) for the major loss of offsite communications capability, offsite response capability, and emergency assessment capability. That reporting criterion is not proposed for removal.

Comment 3 (002-4): The NRC should engage further with communities near commercial nuclear power plants for this rulemaking, including in-person engagement, and perform an environmental justice impact review.

NRC Response: The NRC disagrees with the comment. Although the NRC continuously aims to improve communications, the NRC determined that the elimination of the recommended nonemergency event notifications would not impact local communities because these notifications are not needed for the NRC to fulfill its mission to provide reasonable assurance of adequate protection of the public health and safety, which includes the health and safety of communities around commercial nuclear power plants. To that end, the NRC would retain a number of reporting requirements, including the non-emergency event reports for (1) deviations from technical specifications under § 50.54(x) per § 50.72(b)(1); (2) plant shutdowns required by technical specifications per § 50.72(b)(2)(i); (3) actuations of the reactor protection system per § 50.72(b)(2)(iv)(B); (4) the discovery of degraded or unanalyzed conditions per § 50.72(b)(3)(ii); and (5) the loss of emergency response capabilities per § 50.72(b)(xiii), as those reporting requirements help facilitate the NRC's ability to ensure public safety.

Accordingly, the Commission did not change the NRC staff's recommendations in the regulatory basis in response to this comment.

Comment 4 (002-5, 008-2): The NRC should consider the costs to States to create their own event notification requirements in lieu of those the NRC recommends for deletion.

NRC Response: The NRC agrees with the comments. The NRC incorporated changes to the regulatory analysis to account for States that have indicated that they would initiate procedures to implement any nonemergency event notification requirements at a State level that would be eliminated by this proposed rule. The addition of these costs did not change the overall cost-effectiveness of this proposed rule.

Accordingly, the Commission did not change the NRC staff's recommendations in the regulatory basis in response to these comments.

Comment 5 (004-2): The NRC should be aware that the actual burden hours of evaluating conditions for reportability may exceed the estimated hours assumed in the cost analysis.

NRC Response: The NRC agrees with the comment. The hours assumed in the regulatory analysis are the NRC's best estimate of average hours expended per report. The NRC acknowledges that the actual burden of reports is variable and that some reports may require more time than assumed in the regulatory analysis. However, much of the burden associated with evaluating the conditions would also be needed for licensees' operability evaluations independent of reportability requirements.

Accordingly, the Commission did not change the NRC staff's recommendations in the regulatory basis in response to this comment.

Comment 6 (006-2, 006-03): The NRC should consider ways to accelerate the rulemaking schedule and update existing guidance in NUREG-1022 for reporting degraded or unanalyzed conditions under § 50.72(b)(3)(ii) as soon as practical.

NRC Response: The NRC disagrees, in part, with the comment. The NRC considered accelerating the schedule for this rulemaking but decided not to ( printed page 60708) change the schedule. The rulemaking schedule provides time for proper stakeholder engagement while ensuring that the rule and guidance adequately and appropriately address all issues and public comments. If the NRC proceeds to issue a final rule, the NRC anticipates publication of a final rule in late 2026. Updates to the rulemaking schedule will be made available in the NRC website: www.nrc.gov/​reading-rm/​doc-collections/​rulemaking-ruleforum/​active/​ruleindex.html.

The NRC agrees that updates to existing guidance in NUREG-1022 solely for reporting degraded or unanalyzed conditions under § 50.72(b)(3)(ii) can be achieved in advance of the rulemaking process. The updated guidance would provide additional information to address the high historical retraction rate of degraded or unanalyzed condition notifications under § 50.72(b)(3)(ii) in comparison to the other nonemergency event reporting criteria. Because it is not proposing to amend § 50.72(b)(3)(ii), the NRC can revise the associated guidance independent of this rulemaking effort and on a faster schedule than the rulemaking. To that end, on October 2, 2024, the NRC published in the Federal Register a notice of the issuance of NUREG-1022, Revision 3, “Event Report Guidelines: §§ 50.72 and 50.73,” Supplement 2 (89 FR 80268), to provide clarifications for Section 3.2.4, “Degraded and Unanalyzed Conditions,” of NUREG-1022, Revision 3.

Accordingly, the Commission did not change the NRC staff's recommendations in the regulatory basis in response to these comments.

Comment 7 (008-3): It is unclear from the regulatory basis cost estimates that the rulemaking will provide any significant reduction in reporting burden.

NRC Response: The NRC disagrees with the comment. The regulatory analysis performed for this rulemaking documents the reduction in burden and associated costs resulting from this rulemaking. The analysis shows that the rulemaking is generally cost effective.

Accordingly, the Commission did not change the NRC staff's recommendations in the regulatory basis in response to this comment.

Comment 8 (009-1, 009-2):

The NRC should be careful not to reduce reporting requirements further than prudent.

NRC Response: The NRC agrees with the comments. The regulatory basis used a decision-making process driven by (1) past reporting data, (2) NRC actions in response to those reports, and (3) a qualitative risk assessment based on the NRC's Principles of Good Regulation and the “Be riskSMART” framework, to determine which reporting requirements must be maintained to support the NRC's ability to fulfill its mission.

Accordingly, the Commission did not change the NRC staff's recommendations in the regulatory basis in response to these comments.

Comment 9 (009-3): The NRC should consider additional reporting requirements, such as those related to physical cybersecurity threats, telecommunications incidents, or preparation for and response to severe weather events.

NRC Response: The NRC disagrees with the comment. Although physical and cybersecurity issues are not covered by the § 50.72(b) event reporting requirements, they are subject to reporting to the NRC via other requirements such as those in § 73.77, “Cyber security event notifications,” and 73.1200, “Notifications of physical security events.” Therefore, adding requirements to report these events under § 50.72(b) would be unnecessarily redundant.

Telecommunications issues impacting emergency communications are currently reported to the NRC as required by § 50.72(b)(3)(xiii). Severe weather impacts are not specifically reportable to the NRC via event notifications. However, the NRC maintains general awareness of these impacts through other means, e.g., public meteorology sources, routine inspections, and event reports required by conditions created by the severe weather. Therefore, the NRC did not consider adding a notification requirement for severe weather. Additionally, natural phenomena, such as severe weather, that pose a threat to licensees are reportable via licensee event report (LER) as required by § 50.73(a)(2)(iii).

Accordingly, the Commission did not change the NRC staff's recommendations in the regulatory basis in response to this comment.

Comment 10 (009-4): Eliminating the event notification reporting requirements could result in licensees eliminating their procedures to voluntarily notify resident inspectors of issues the same day. Also, it would require resident inspectors to track due dates for licensee event reports, identify when they are not provided within 60 days, and follow up on non-reported items via the corrective action program, which is an additional burden contrary to the Commission's directive.

NRC Response: The NRC disagrees with the comment. Although licensees often notify resident inspectors of event notifications, any internal licensee requirement to do so may be changed regardless of this rulemaking, as there is no existing regulatory requirement for them to notify resident inspectors. Resident inspectors have numerous other means of maintaining awareness of plant activities and conditions, which would not be impacted by this rulemaking.

Additionally, resident inspectors sample plant condition reports to ensure licensees submit required notifications, including event notifications and licensee event reports. This rulemaking would not impact that activity.

Finally, resident inspector follow-up of events happens largely in parallel with the development of event notification reports, not solely in response to event notification reports. The removal of event notifications would not substantially change the scope of resident inspector activities. Rather, it means that plant conditions meeting the criteria recommended for elimination under this rulemaking would no longer require specific follow-up due to an event notification report. Instead, such conditions would be treated as any other routine condition of potential interest to resident inspectors, and any follow-up activities inspectors choose to take would be driven via routine baseline activities.

Accordingly, the Commission did not change the NRC staff's recommendations in the regulatory basis in response to this comment.

Comment 11 (009-5, 009-6): The NRC should not eliminate both the 4-hour reporting requirement to report discharge of the emergency core cooling system (ECCS) in § 50.72(b)(2)(iv)(A) and the requirement to report safety system actuation in § 50.72(b)(3)(iv), because then there will be no notification of these types of events.

NRC Response: The NRC disagrees with the comments. In Section 14.0 of the regulatory basis, the NRC staff determined that the 4-hour event notification requirement in § 50.72(b)(2)(iv)(A) is unnecessary. Nearly all events resulting in ECCS discharge into the reactor coolant system that are reported to the NRC are captured by the § 50.72(b)(2)(iv)(B) requirement to report reactor protection system (RPS) actuations because most ECCS actuations are also RPS actuation signals. The NRC's analysis determined that situations where an ECCS actuates or is required to actuate, but a corresponding RPS actuation is not required, are not significant enough to ( printed page 60709) warrant prompt NRC actions. Risk-significant ECCS actuations result in corresponding RPS actuations when the reactor is at power. There are few risk-significant conditions where ECCS actuations may occur and RPS is not required to be active. Of those, such as during refueling activities with the reactor coolant system partially drained, the potential significance of the conditions is more appropriately captured by licensees' emergency action levels (EALs) and the associated notifications required in licensees' emergency plans. For non-risk-significant conditions where an ECCS actuation occurs and RPS actuation does not also occur, the NRC would not need to take prompt actions to fulfill its mission. These conditions are appropriately captured by the corresponding 60-day LER reporting requirements and routine baseline inspection activities.

Separately, in Section 19.0 of the regulatory basis, the NRC considered the 8-hour ECCS actuation reporting requirement as a subset of the reporting requirements for general safety system actuations in § 50.72(b)(3)(iv). The NRC found that prompt NRC actions were not warranted. For conditions reportable under this criterion (including general ECCS actuation signals, but not ECCS discharges that also result in RPS actuations), the NRC can fulfill its mission through routine inspections and the corresponding 60-day LER reporting requirements.

Accordingly, the Commission did not change the NRC staff's recommendations in the regulatory basis in response to these comments.

Comment 12 (010-1): The NRC should make the conditions reportable under § 50.72(b)(3)(xiii) reportable as an LER.

NRC Response: The NRC disagrees with the comment. As discussed in the “Licensee Event Report System” final rule (48 FR 33850, July 26, 1983), when the NRC established the LER system, the purpose of reports required via § 50.73 are for the NRC to obtain detailed narratives of potentially safety significant events. In the review of past reports made under this criterion, the NRC did not find any reports where the conditions requiring an event notification under § 50.72(b)(3)(xiii)—yet falling below the threshold of being reportable under other criteria—would be of sufficient potential safety significance to warrant the need for a detailed description through the LER reporting of § 50.73. As such, NRC awareness of these events would be adequately maintained under the NRC's baseline inspection program.

Accordingly, the Commission did not change the NRC staff's recommendations in the regulatory basis in response to this comment.

Comment 13 (012-1): The proposed changes may increase the burden on resident inspectors and require resident inspectors to devote more inspection time to plant status.

NRC Response: The NRC disagrees with the comment. Resident inspectors maintain awareness of a broad array of plant activities and conditions. Most of these conditions are not reportable. The elimination of the proposed reporting requirements does not imply that residents should track those conditions outside of event reports. Residents would still maintain awareness of plant activities and conditions via their normal methods, and residents would not have to give any more or less attention to those that were previously reportable than they would give to other plant conditions of interest that were not previously reportable.

Accordingly, the Commission did not change the NRC staff's recommendations in the regulatory basis in response to this comment.

Comment 14 (012-4): There appears to be a gap in the current requirements. The NRC requires licensees to report a loss of outbound (licensee to NRC) communication capability but does not require licensees to report a loss of inbound (NRC to licensee) communication capability. Section 50.72 should be modified to require notification of the NRC for any loss of communications capability (inbound or outbound).

NRC Response: The NRC agrees, in part, with this comment. The current reporting requirements do not require licensees to notify the NRC of a loss of inbound communications capability. Licensees are not required to report a loss of inbound communications because there is no reliable way for licensees to test or otherwise know that inbound communications capabilities are lost. However, the NRC tests inbound communications daily with routine calls between licensees and the NRC's Headquarters Operations Officers. This testing allows the NRC to become aware of any loss of inbound communications capabilities.

Accordingly, the Commission did not change the NRC staff's recommendations in the regulatory basis in response to this comment.

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

The NRC is proposing revisions to decommissioning licensing regulations in 10 CFR parts 20, 50, 52, and 53. Proposed changes to decommissioning requirements build on prior NRC initiatives, including the rulemaking effort discussed in SECY-24-0011, “Regulatory Improvements for Production and Utilization Facilities Transitioning to Decommissioning.” However, given that the Commission is presently considering the NRC staff recommendations in SECY-24-0011, the revisions in this proposed rule would be separate from, and would not rely on, those recommendations.

A. Expansion of Partial Site Release Requests

In 2003, the NRC published a final rule that added § 50.83, “Release of part of a power reactor facility or site for unrestricted use” (68 FR 19711). Section 50.83 requires NRC approval for a partial site release for unrestricted use at a reactor site before NRC approval of the licensee's License Termination Plan (LTP). Before that final rule, there was no guidance for allowing a portion of a site to be released for unrestricted use, and the entire license footprint must have been decommissioned at the end of the facility's operating period. The flexibility of a partial site release allowed 10 CFR part 50 power reactor licensees to request release of portions of the facility at any time prior to the approval of the LTP, including during facility operations. Partial releases following LTP approval would be governed by the LTP or changes thereto, and partial releases for restricted use are not permitted prior to LTP approval. Under § 50.83, licensees may request a partial site release by providing specific information to the NRC, where the extent of the information necessary depends on whether the area to be released has been designated as “nonimpacted” or “impacted.”

In 2007, the NRC published a final rule amending 10 CFR part 52 (72 FR 49517). That final rule included in 10 CFR part 52 provisions analogous to the decommissioning and license termination sections in 10 CFR part 50, yet the 10 CFR part 52 provisions reference only §§  50.51 and 50.82, which do not include requests for partial site release. Part 53 also includes a partial site release provision in § 53.1080, “Release of part of a commercial nuclear plant or site for unrestricted use.” This proposed rule would correct this exclusion in 10 CFR part 52 by including a new section, § 52.111, “Release of part of a power reactor facility or site for unrestricted use,” to add a specific provision for release of part of a site for unrestricted use before license termination for 10 ( printed page 60710) CFR part 52 licensees. The language in proposed § 52.111 is nearly identical to the language in § 50.83, with specific applicability to holders of a combined license under 10 CFR part 52. The NRC is also proposing minor conforming changes to 10 CFR part 20, “Standards for Protection Against Radiation,” subpart E, “Radiological Criteria for License Termination,” to include the release of part of a facility or site under the proposed § 52.111 and to limit the circumstances for which the Commission could require additional cleanup following release for unrestricted use. The proposed changes would provide 10 CFR part 52 licensees with a clear and efficient path toward releasing a portion of their site prior to license termination and would align the regulations in 10 CFR part 52 with those in 10 CFR parts 50 and 53, without any reduction in safety or security.

B. Reducing Regulatory Requirements for Prompt Decommissioning of Power Reactors

In 1996, the NRC amended its regulations to require power reactors licensed under 10 CFR part 50 to submit a Post-Shutdown Decommissioning Activities Report (PSDAR) before or within 2 years of permanent cessation of operations instead of a detailed Decommissioning Plan (DP) (61 FR 39278). Prior to the 1996 amendment, the DP approach, which is currently the required licensing mechanism for nonpower production or utilization facilities and all nuclear materials facilities, prohibited power reactor licensees from performing dismantlement activities until the NRC reviewed and approved the DP and issued an associated order permitting decommissioning activities in accordance with the approved plan.

Given that power reactor licensees have 60 years to terminate their license per §§ 50.82(a)(3), 52.110(c), and 53.1070(c), the introduction of the PSDAR allowed these licensees to either begin dismantlement activities promptly or place the reactor in storage and delay dismantlement, the standard industry practice at the time. During the development of the 1996 rule, the NRC assumed that power reactor licensees would not be ready to terminate their license within a few years after permanent cessation of operations. The PSDAR does not require NRC approval, rather, under §§ 50.82(a)(9)(i), 52.110(i)(1), and 53.1070(i)(1), the licensee must submit the LTP for NRC review and approval before 2 years of expected license termination. Decommissioning must be performed in accordance with the approved LTP.

Under current regulations in §§ 50.82(a)(4)(i), 52.110(d)(1), and 53.1070(d)(1), the PSDAR submitted to the NRC must contain a schedule of planned decommissioning activities, an estimate of the decommissioning costs expected to be incurred, and a discussion of environmental impacts of decommissioning. Under current regulations in §§ 50.82(a)(9)(ii), 52.110(i)(2), and 53.1070(i)(2), the LTP submitted to the NRC must contain identification of remaining dismantlement activities, an updated site-specific estimate of remaining decommissioning costs, and supplement to the environmental report describing any new information or significant environmental change associated with the licensee's proposed termination activities. In addition, the LTP must include a site characterization; plans for site remediation; detailed plans for the final radiation survey; a description of the end use of the site, if restricted; and identification of parts, if any, of the facility or site that were released for use before approval of the LTP. The NRC recognizes the overlap in information submitted in both the PSDAR and LTP. If a licensee planned to promptly terminate the license after permanent shutdown, much of the information would be redundant.

The future landscape of power reactors licensed under 10 CFR parts 50, 52, and 53 will include various reactor technologies, types, and sizes. Therefore, the decommissioning of such reactors will vary tremendously, and accordingly, the NRC is proposing to proactively amend the decommissioning licensing process to provide regulatory enhancements and flexibility in the decommissioning and subsequent license termination of all potential reactors. To accommodate licensees who may decommission and terminate their license promptly after shutdown, the NRC is proposing to eliminate the requirement to submit a PSDAR if an LTP is submitted before or within 2 years of permanent cessation of operations.

To implement this proposed change, the NRC proposes conforming changes to certain requirements that commence at the time the PSDAR is submitted in §§ 50.82(a)(5), 50.82(a)(7), 52.110(e), 52.110(g), 53.1070(e), and 53.1070(g). The NRC is also proposing to amend the duplicative requirement in §§ 50.82(a)(8)(iii) and 52.110(h)(3) related to submission of a site-specific decommissioning cost estimate.

C. Public Engagements Related to Decommissioning Licensing Actions

Under current NRC regulations §§ 50.82(a)(4)(ii), 52.110(d)(2), and 53.1070(d)(2), the NRC must notice receipt of the PSDAR in the Federal Register , make the PSDAR available for public comment, hold a public meeting in the vicinity of the site being decommissioned, and publish details about the public meeting in the Federal Register and in a local forum. In the nearly 30 years since the creation of the PSDAR, the NRC has gained experience with preparing and reviewing PSDARs and conducting associated public interactions. Public engagements related to submittal of a PSDAR impose significant costs on the NRC and licensees without commensurate safety benefits, and public comments have rarely resulted in changes to a licensee's PSDAR. This proposed rule would eliminate the requirement for the NRC to notice receipt of the PSDAR and make the PSDAR available for public comment. The NRC would continue to use the information in the PSDAR to maintain public health and safety at the beginning of the decommissioning process.

Additionally, this proposed rule would modify the locality requirement for the NRC to hold a public meeting for the PSDAR, LTP, or a partial site release request “in the vicinity of the site,” as required by §§ 50.82(a)(4)(ii), 50.82(a)(9)(iii), 50.83(f), 52.110(d)(2), 52.110(i)(3), 52.111(f), 53.1070(d)(2), 53.1070(i)(3), and 53.1080(f). This proposed rule would maintain the requirement for the NRC to host a public meeting and modernize NRC regulations to accommodate current and future practices for how the NRC currently conducts public meetings in support of its long-standing position that public involvement in NRC activities is a cornerstone of strong, fair regulation of the nuclear industry.

D. Shortening the Period After Submittal of a PSDAR Before Major Decommissioning Activities May Begin

Under current NRC regulations in §§ 50.82(a)(5), 52.110(e), and 53.1070(e), a power reactor licensee may not begin major decommissioning activities until 90 days after the NRC receives the PSDAR and until certifications of permanent cessation of operations and permanent removal of fuel from the reactor vessel have been submitted. The purpose of the current 90-day restriction was for the NRC to engage stakeholders during its review of the PSDAR before dismantlement and remediation began onsite. Because this proposed rule ( printed page 60711) would eliminate the required public comment period associated with the PSDAR, the NRC's review of future PSDAR submittals would be completed much faster. The NRC anticipates these reviews would be completed within 30 days. Therefore, this proposed rule would reduce the timeframe from 90 days to 30 days after the NRC receives the submittal required before or within 2 years of permanent cessation of operations ( i.e., PSDAR or LTP).

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

A. Conforming the Financial Qualifications Requirements for 10 CFR Parts 50 and 52 Applicants With the Financial Qualifications Requirements for 10 CFR Part 53 Applicants

Under §§ 50.33(f) and 52.77, applications for utilization facility permits and licenses under 10 CFR parts 50 and 52, respectively, are required to include, except for an electric utility applicant for a license to operate a utilization facility of the type described in paragraph (b) of § 50.21, “Class 104 licenses; for medical therapy and research and development facilities,” or in § 50.22, “Class 103 licenses; for commercial and industrial facilities,” information sufficient to demonstrate that the applicant “possesses or has reasonable assurance of obtaining” the funds necessary for construction and/or operation, as applicable, of the facility along with associated financing details. However, for commercial nuclear plants licensed under 10 CFR part 53, the NRC uses the less restrictive financial qualifications standard of “appears to be financially qualified,” which is similar to the standard used in paragraph (a)(5) of § 70.23, “Requirements for the approval of applications,” for the possession of special nuclear material.

In 2018, the NRC staff submitted to the Commission a draft proposed rule to change the NRC's financial qualifications requirements in 10 CFR part 50 to match the language of 10 CFR part 70. In 2022, the Commission issued SRM-SECY-18-0026, “Proposed Rule: Financial Qualifications Requirements for Reactor Licensing,” disapproving the draft proposed rule. Instead, the Commission directed the staff to address this issue during the development of 10 CFR part 53 and to solicit stakeholder feedback on financial qualifications. As directed, the staff solicited stakeholder feedback on using the 10 CFR part 70 standard in 10 CFR part 53 and, based on that feedback, 10 CFR part 53 ultimately did use the 10 CFR part 70 standard.

The NRC is now proposing to change § 50.33(f)(1) and (2) so that they conform to the financial qualifications standard for 10 CFR part 53 applicants. Specifically, the proposed changes to § 50.33(f)(1) and (2) would replace the current “possesses or has reasonable assurance of obtaining the funds necessary” financial qualifications standard with the 10 CFR part 53 “appears to be financially qualified” financial qualifications standard. And, as with 10 CFR part 53, the proposed changes would provide that if available funding at the time of application is 50 percent or less, the applicant should include proposed license conditions to facilitate verification that funding is available prior to the start of construction and/or operation, as applicable. These changes would ensure that the NRC's regulations are consistent.

B. Decommissioning Trust Fund Use

In 2024, the NRC issued “Interim Staff Guidance: Use of the Decommissioning Trust Fund During Operations for Major Radioactive Component Disposal” (89 FR 63447), providing both the NRC's position on the topic and information that would assist the NRC staff in evaluating a request for an exemption from § 50.82 to allow for a decommissioning trust fund (DTF) withdrawal for major radioactive component disposal during operations. To date, the NRC has evaluated and approved two separate exemption requests related to the use of the DTF during operations and is currently evaluating a third. The NRC expects a significant number of similar requests in the foreseeable future. The NRC evaluates each exemption request for a DTF withdrawal during operations to ensure that it meets the criteria for exemptions in § 50.12, “Specific exemptions,” and considers the totality of facts in determining whether to grant or deny the request. Each request is accompanied by a safety and environmental review, which results in an average review time of 9 months.

To reduce time and resource burdens on both the industry and the NRC, rather than relying on the exemption process, the NRC proposes to establish a 60-working day notice/negative consent process. This proposed rule changes would explicitly allow the use of DTFs during operations for radiological decommissioning activities, including, but not limited to, major radioactive component disposal, under certain defined circumstances. This proposed rule would not change the NRC's existing practice of evaluating a DTF withdrawal notice for “extraordinary circumstances,” using guidance similar to the current Interim Staff Guidance. Additionally, proposed changes to §§ 50.75 and 53.1045 would maintain the condition that licensees must continue to have sufficient funds available for decommissioning in the DTF. More information is provided in the draft Regulatory Guide (DG)-1476 (proposed RG 1.159, Revision 4), “Assuring the Availability of Funds for Decommissioning Nuclear Reactors.”

The NRC would continue to implement its robust decommissioning funding assurance oversight program, including the review of decommissioning funding status reports for operating reactors. These reports enable the NRC to ensure that funding will be available to safely decommission a facility that has permanently ceased operation, and to identify and respond to shortfalls in reasonable assurance, should they occur. These activities would provide an appropriate level of oversight to prevent funding shortages.

10 CFR Part 50—Risk Informed Seismic Design

Within the current regulatory framework, all power reactor applicants under 10 CFR parts 50 and 52 are required to meet appendix S, “Earthquake Engineering Criteria for Nuclear Power Plants,” to 10 CFR part 50, which does not allow for a graded approach to seismic design based on the risks of the facility or the SSCs within the facility. Appendix S to 10 CFR part 50 currently only allows for use of a single design-basis ground motion (DBGM), referred to as the safe-shutdown earthquake ground motion (SSE), and imposes a minimum peak ground acceleration (consistent with the highest risk level) for the response spectrum at the foundation level of structures.

Recently developed guidance, DG-4036 (proposed RG 4.29), “Graded Approach to Site Characterization for New Reactor Applications,” which would support changes to 10 CFR part 100, “Reactor Site Criteria,” in the proposed rule, “Modernizing Reactor Licensing, Safety Oversight, and Siting Practices” (91 FR 44560), would provide guidance on determining appropriate DBGMs for a facility based on the potential consequences from a design basis seismic event. DG-4036 also would include an appendix providing guidance on optimization of a facility's seismic design basis based on the use of multiple seismic design categories and associated DBGMs. This proposed rule would modify appendix S to 10 CFR part 50 to allow use of these risk-informed approaches under 10 CFR ( printed page 60712) parts 50 and 52. A guidance document supporting this proposed rule, DG-1471 (Draft RG 1.208 Revision 2), “A Performance-Based Approach to Define the Site-Specific Earthquake Ground Motion,” would rely on the proposed changes to 10 CFR part 100.

Section 53.480 includes a risk-informed and performance-based alternative to appendix S to 10 CFR part 50 that allows for the use of multiple DBGMs for 10 CFR part 53 applicants and licensees. The different target performance goals associated with each DBGM provide varying levels of reliability during a seismic event, which can be assigned based on the safety-significance of SSCs or potential radiological consequence resulting from failure of the SSCs during a seismic event and enable the gradation of seismic design criteria for SSCs. Similarly, and in a separate rulemaking, the proposed 10 CFR part 57, “Licensing Requirements for Microreactors and Other Reactors with Comparable Risk Profiles” (91 FR 23628), would provide a risk-informed and performance-based regulatory framework that would not preclude the use of multiple DBGMs for 10 CFR part 57 applicants and licensees. The proposed update to appendix S to 10 CFR part 50 in this rulemaking would be consistent with the risk-informed and performance-based approaches addressed in these other regulatory frameworks and DG-4036 and would enable the implementation of such approaches under 10 CFR parts 50 and 52.

These proposed rules and guidance would enable nuclear facilities to avoid overly conservative seismic design requirements by risk informing seismic design to the appropriate level needed to ensure safety.

10 CFR Part 50—Probabilistic Risk Assessment Requirements

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

The NRC is proposing to amend § 50.69 to extend the applicability of its alternative approach to holders of an LWR CP under 10 CFR part 50, as well as to applicants for an LWR design certification and holders of an LWR COL and ML under 10 CFR part 52.

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

The NRC regulations include several backfitting and issue finality rules that provide for the predictability and stability of NRC regulations. Although these rules vary to an extent, each generally requires the NRC to demonstrate that specific criteria are met before imposing new or revised requirements on certain holders of, and certain applicants for, NRC approvals.

The NRC has always considered nuclear power reactors licensed under section 103 or 104b of the Atomic Energy Act of 1954, as amended (AEA) ( i.e., commercial nuclear power reactors) to be within the scope of § 50.109, “Backfitting.” In 2026 (91 FR 2287), the NRC issued a notice of interpretation explaining that commercial non-power production or utilization facilities (NPUFs) licensed under section 103 of the AEA are within the scope of § 50.109 and non-commercial NPUFs licensed under section 104 of the AEA are excluded from the scope of § 50.109. This proposed rule would revise § 50.109 to clarify this applicability.

The current § 50.109 does not require the NRC to consider the costs of its actions in cases of ensuring, defining, or redefining adequate protection, or in cases of ensuring compliance with NRC requirements or conformance with written licensee commitments. Unless a backfit is necessary for ensuring, defining, or redefining adequate protection or is necessary for compliance, § 50.109 requires the NRC to demonstrate through a systematic and documented analysis ( i.e., a backfit analysis) that there is a substantial increase in the overall protection of public health and safety or the common defense and security to be derived from the backfit and that the direct and indirect costs of implementation are justified in view of this increased protection.

The NRC also has backfitting rules in § 70.76, “Backfitting,” for entities licensed to possess special nuclear material in quantities greater than a critical mass and engage in specific activities ( e.g., fuel facilities), and § 72.62, “Backfitting,” for independent spent fuel storage installations and monitored retrievable storage installations. These regulations provide definitions of “backfitting” and related requirements similar to those in § 50.109.

The regulations in 10 CFR part 52 include several issue finality provisions related to the licensing of nuclear power reactors involving early site permits, standard design certifications, COLs, SDAs, and MLs. The design certification rules in the appendices to 10 CFR part 52 also include specific backfitting and issue finality provisions. The issue finality provisions generally accomplish the same objective as the backfitting provisions by limiting changes to certain decisions made as part of the 10 CFR part 52 licensing process. For example, there are several issue finality provisions in 10 CFR part 52 that require justification similar to § 50.109 for NRC-initiated changes affecting issue finality.

The regulations in 10 CFR part 53 also include a backfitting rule in § 53.1590, “Backfitting,” and issue finality provisions related to the licensing of nuclear power reactors under 10 CFR part 53. The backfitting rule and issue finality provisions in 10 CFR part 53 are generally equivalent to the backfitting rule in § 50.109 and issue finality provisions in 10 CFR part 52.

Backfitting generally does not include changes made by the NRC at the request of an applicant, licensee, or similar entity, such as changes made in response to a request for a license amendment, exemption, or departure. Changes affecting issue finality can be initiated by an applicant, licensee, or the NRC.

On November 29, 2016, the Commission issued SRM-COMSECY-16-0020, “Revision of Guidance Concerning Consideration of Cost and Applicability of Compliance Exception to Backfit Rule” (ML16334A462), and directed the staff to consider costs in compliance backfit justifications and to make other changes in how the backfit and issue finality rules are implemented. The term “compliance backfit” encompasses both backfitting actions and changes affecting issue finality to ensure compliance with NRC requirements. This direction was based, in part, on the Supreme Court's 2015 decision in Michigan v. Environmental Protection Agency, 576 U.S. 743 (2105), that reflects the view that under the Administrative Procedure Act, unless Congress has indicated otherwise, an agency's decision-making calculus should include at least some consideration of the cost placed on a ( printed page 60713) licensee to comply with new requirements.

Under the NRC's current regulations, the NRC uses a compliance backfit justification when a modification is necessary to bring a facility into compliance with NRC requirements or into conformance with written commitments by the licensee. Compliance backfitting is limited to addressing situations in which the NRC approved a method of compliance with a requirement and later determines that (1) the method does not comply with the requirement and (2) the previous approval occurred because of an omission or mistake of fact at the time of approval. New or modified interpretations of what constitutes compliance are not within the scope of the compliance backfit justification but may be considered under the other applicable justifications for backfitting or imposing a change affecting issue finality. Prior to SRM-COMSECY-16-0020, the NRC evaluation of compliance and adequate protection backfits or changes affecting issue finality were similar, and costs were not considered in either case.

On September 19, 2019, NRC Management Directive (MD) 8.4, “Management of Backfitting, Forward Fitting, Issue Finality, and Information Requests,” was revised to reflect the Commission's updated policy in SRM-COMSECY-16-0020 on the implementation of the backfitting and issue finality rules. As described in the revised MD 8.4, if the NRC is considering a backfitting action or change affecting issue finality, it must first consider whether regulatory action is an issue of adequate protection ( i.e., necessary to provide reasonable assurance of adequate protection or involving defining or redefining adequate protection) and, if so, whether there is an imminent threat to public health and safety. If regulatory action is not an issue of adequate protection, then the NRC needs to determine if the proposed action satisfies the applicable compliance backfit criteria, which includes cost considerations. Finally, if the proposed action does not satisfy the applicable compliance backfit criteria, then the NRC may consider whether the proposed action is a cost-justified substantial increase in overall protection. The issue finality provisions can vary from this approach because some issue finality provisions allow for other justifications and some do not allow for a cost-justified substantial increase in overall protection.

Therefore, by the process described in MD 8.4, a compliance backfit is not necessary for reasonable assurance of adequate protection of public health and safety or the common defense and security. The NRC must consider costs in the compliance backfit justification but, in contrast to a cost-justified substantial increase in overall protection, the NRC is afforded substantial flexibility in its cost considerations and does not need to demonstrate a substantial increase in overall protection.

On June 6, 2022, the NRC staff requested Commission approval to publish the proposed rule described in SECY-22-0052, “Proposed Rule: Alignment of Licensing Processes and Lessons Learned from New Reactor Licensing (RIN 3150-Al66)” (package ML21159A055). In response, the Commission issued SRM-SECY-22-0052 (ML24326A003), which approved the publication of a revised proposed rule that would, among other things, clarify the backfitting and issue finality requirements in 10 CFR parts 50 and 52 that govern changes to early site permits, SDAs, and manufacturing licenses. These proposed clarifications and other proposed changes are included in this rulemaking with some differences to what was described in SECY-22-0052.

E.O. 14300, section 5(f), directs the NRC to “[e]stablish stringent thresholds for circumstances in which the NRC may demand changes to reactor design once construction is underway” as part of the wholesale revision of its rules and guidance documents. The NRC's backfitting regulations in §§ 50.109 and 53.1590 and issue finality provisions in 10 CFR parts 52 and 53, among other regulations, include such thresholds. Additionally, in response to E.O. 14267, the Commission identified that several backfitting and issue finality provisions could be modified to reduce their competitive burden. In response to E.O.s 14300 and 14267, the NRC proposes to amend its backfitting and issue finality rules to eliminate the compliance backfit exception, revise and clarify the applicability of the backfitting and issue finality requirements, add criteria to allow the NRC to reduce regulatory burden, and make other corrections.

Specifically, the proposed amendments would revise the backfitting rules in §§ 50.109, 53.1590, 70.76, and 72.62. This proposed rule does not apply to the backfitting rule in 10 CFR part 76, “Certification of Gaseous Diffusion Plants,” because 10 CFR part 76 is no longer used and is scheduled to be sunset on January 8, 2027. The proposed amendments would also revise the following issue finality rules that apply to early site permits, design certifications, and manufacturing licenses issued under 10 CFR parts 52 or 53: §§ 52.31, “Criteria for renewal,” 52.39, “Finality of early site permit determinations,” 52.59, “Criteria for renewal,” 52.63, “Finality of standard design certifications,” 52.171, “Finality of manufacturing licenses; information requests,” 52.179, “Criteria for renewal,” 53.1176, “Criteria for renewal,” 53.1188, “Finality of early site permit determinations,” 53.1257, “Criteria for renewal,” 53.1263, “Finality of standard design certifications,” 53.1288, “Finality of manufacturing licenses,” and 53.1295, “Renewal of manufacturing licenses.”

In addition, the proposed amendments would revise the plant-specific backfitting limitations in paragraph VIII.B.3 of the five design certification rules currently in effect:

a. Appendix A to 10 CFR part 52, “Design Certification Rule for the U.S. Advanced Boiling Water Reactor,”

b. Appendix D to 10 CFR part 52, “Design Certification Rule for the AP1000 Design,”

c. Appendix E to 10 CFR part 52, “Design Certification Rule for the ESBWR Design,”

d. Appendix F to 10 CFR part 52, “Design Certification Rule for the APR1400 Design,” and

e. Appendix G to 10 CFR part 52, “Design Certification Rule for NuScale.”

Two design certifications have expired because no timely renewal applications were submitted in accordance with § 52.57, “Application for renewal.” These are appendix B to 10 CFR part 52, “Design Certification Rule for the System 80 + Design,” and appendix C to 10 CFR part 52, “Design Certification Rule for the AP600 Design.” Accordingly, this proposed rule would apply to the five design certifications currently in effect and would not apply to the two expired design certifications.

Parts 52 and 53—Notification of Initial Fuel Load

The current regulations in §§ 52.103(a) and 53.1452(a) require an applicable licensee to notify the NRC of the licensee's scheduled date for initial loading of fuel at least 270 days before the scheduled date and subsequently to notify the NRC of any updates to the schedule every 30 days thereafter. In those cases where, after providing the initial notification, the licensee experiences extended delays, the licensee would still be required to provide updates every 30 days even if the schedule no longer changes. The regulations in § 53.1452(a) also require that COL holders installing fueled manufactured reactors notify the NRC at ( printed page 60714) least 270 days prior to initiating the removal of the features to prevent criticality and subsequently to notify the NRC of any updates to the schedule every 30 days thereafter. This proposed rule would eliminate the requirement for recurrent reporting of updates to the licensee's scheduled date for initial fuel load every 30 days after the initial report submitted no less than 270 days before the scheduled date. Instead, this would be replaced with a requirement to notify the NRC only in those cases where, after providing the initial notification, the scheduled date for initial fuel load is revised.

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

The regulations in §§ 52.47(a)(21) and (22), 52.79(a)(20) and (37), 52.137(a)(21) and (22), and 52.157(f)(28) and (29) require applicants to submit, for NRC review, certain generic issues and operating experience from generic communications technically relevant to the associated design. As explained in Regulatory Guide 1.206, “Applications for Nuclear Power Plants,” Revision 1, October 2018, the scope of the generic issues to be evaluated is those Unresolved Safety Issues (USIs) and medium- and high-priority Generic Safety Issues (GSIs) that (1) are identified in the version of NUREG-0933, “A Prioritization of Generic Safety Issues,” current on the date up to 6 months before the docket date of the application and (2) are technically relevant to the design. RG 1.206, Revision 0, “Combined License Applications for Nuclear Power Plants,” also states that the scope of the operating experience from generic communications to be evaluated is those Generic Letters and Bulletins issued after the most recent revision of NUREG-0800, “Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition,” and 6 months before the docket date of the application, or comparable international operating experience.

The Commission's Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants (50 FR 32138; August 8, 1985) and Policy Statement on Standardization (52 FR 34884; September 15, 1987) introduced the policy of evaluating USIs and medium- and high-priority GSIs, including a special focus on assuring the reliability of decay heat removal systems and the reliability of electrical supply systems. The 1989 10 CFR part 52 rulemaking (54 FR 15372; April 18, 1989) codified the policy, and the subsequent 10 CFR part 52 rulemaking (72 FR 49352; August 28, 2007) introduced the provisions related to operating experience.

As discussed in SECY-15-0002, “Proposed Updates of Licensing Policies, Rules, and Guidance for Future New Reactor Applications” (ML13277A420), while 10 CFR part 50 does not contain regulations identical to those in 10 CFR part 52, under Commission policy, applicants under 10 CFR part 50 demonstrate the resolution of USIs and GSIs prior to issuance of an operating license.

The NRC proposes to remove this codified requirement for 10 CFR part 52 applicants, and to ensure consistency across frameworks, applicants under 10 CFR part 50 would not be required to demonstrate the resolution of USIs and GSIs prior to issuance of an operating license as well. These applicants would still be required to demonstrate that applicable NRC safety requirements are satisfied. Part 53 does not include requirements related to the evaluation of the generic issues and operating experience as does 10 CFR part 52.

The evaluation of the generic issues and operating experience from generic communications was intended to lead to improved efficiency and effectiveness of the NRC safety reviews. However, experience with new reactor licensing has demonstrated that these requirements have led to applicants expending significant resources without significant contributions to either the effectiveness or efficiency of the associated safety review. Accordingly, the evaluation is an unnecessary burden given the limited benefit. This assessment result is largely based on the following.

First, as the subject generic issues and generic communications were resolved, regulations and guidance have typically been revised or developed to inform future safety reviews. The regulatory guidance, e.g., the Standard Review Plan and Advanced Reactor Content of Application Project (ARCAP) guidance ( www.nrc.gov/​reactors/​new-reactors/​advanced/​modernizing/​guidance/​advanced-reactor-content-of-application-project.html), regulatory guides, and endorsed industry consensus standards, contain provisions and acceptance criteria that reflect the lessons learned from past operating experience including generic issues and generic communications. In addition, as the NRC's approach to new and advanced reactor licensing has substantially evolved, the consideration of operating experience, generally broader than those from generic issues and generic communications, are further integrated into both NRC and industry guidance and practices. An example is the advancement and increased use of probabilistic risk assessment (PRA), which is required under 10 CFR part 52. The industry PRA standard, endorsed by the NRC in regulatory guides, involves an evaluation of operating experience for various technical elements including initiating event analysis and event sequence analysis with an emphasis on a systematic and comprehensive assessment of hazards and associated licensing events of potential risk or safety significance. Accordingly, licensing experience has demonstrated that a separate evaluation of the subject generic issues and generic communications was not necessary as it largely became redundant.

Next, most past generic issues and generic communications are based on the operating LWRs. Their applicability and technical relevance to new and advanced reactors, such as small modular reactors and non-LWR reactors, are very limited. These new and advanced reactor designs, often much simpler and safer using inherent features and passive systems, have eliminated or reduced the potential risk that, historically, was addressed via evaluation of operating experience, including those identified in the generic issues or generic communications. The introduction of advanced reactor designs, such as microreactors, is expected to further limit the applicability and technical relevance of the subject generic issues and generic communications. Further, new and advanced reactor designs are significantly different from the operating reactors and reflect the vast operating experience from the design and operation of nuclear reactors of earlier generations. For example, new and advanced reactors provide enhanced margins of safety and/or use simplified, inherent, passive, or other innovative means to accomplish their safety functions, consistent with the Commission's Policy Statement on the Regulation of Advanced Reactors (73 FR 60612; October 14, 2008).

Finally, since its inception in 1976, the NRC's Generic Issues Program addresses potential safety or security concerns that affect multiple nuclear facilities and are not adequately covered by existing regulations or programs. Most of the generic issues have been identified during the plant operations. USIs are a specific subset of generic issues that are considered to have potential safety significance and require further analysis before a regulatory decision can be made. Over the years, fewer generic issues have been ( printed page 60715) identified, with the most recent issue (GSI-206: Flooding of Nuclear Power Plant Sites Following Upstream Dam Failures) being identified in 2011. Currently, all generic issues have been resolved, generally meaning that regulations or programs are in place for future licensing reviews. Similarly for generic communications, the issuance of generic letters and bulletins has decreased over the years, with the most recently issued Bulletin being in 2012 and Generic Letter in 2016. With sections of NUREG-0800 being updated occasionally, those Bulletins and Generic Letters issued since the last update of NUREG-0800 are minimal or none, especially those that may be technically relevant to new and advanced reactors. The underlying issues associated with Generic Letters and Bulletins, which often involves an NRC information request from the affected licensees, are typically resolved as part of the process implementation of NRC MD 8.18, “Generic Communication Program.”

Accordingly, this rulemaking would eliminate the regulations under 10 CFR part 52 that require applicants to submit for NRC review an evaluation of generic issues and operating experience from generic communications technically relevant to their design.

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

Under current regulations, §§ 52.147 and 53.1218, SDAs expire 15 years after issuance. This expiration period is inconsistent with the duration of other approvals, such as design certifications and combined licenses, that are also supported by renewal provisions.

Moreover, SDAs are currently excluded from the change control processes outlined in §§ 50.59, “Changes, tests, and experiments,” and 53.1550, “Evaluating changes to facility as described in Final Safety Analysis Reports.” They also lack a mechanism for amendments requiring prior NRC approval, limiting the ability of SDA holders to update their designs without initiating a new approval process.

Accordingly, this rulemaking addresses stakeholder feedback and supports the NRC's strategic objectives to modernize its regulatory framework, reduce unnecessary regulatory burden, and foster innovation in reactor design by amending 10 CFR 52.147 and 53.1218 to remove the expiration date for SDAs.

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

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

10 CFR Part 50—Post-TMI Requirements

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

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

A. Minimum Number of Manipulations

In 10 CFR part 55, “Operators' Licenses,” § 55.31(a)(5) currently requires that operator license applicants, as trainees, provide evidence of successfully manipulating the controls of either the plant for which the license is sought or a simulator. It further specifies that at least five significant control manipulations must be performed that affect reactivity or power level.

The NRC identified § 55.31(a)(5) as a regulation that should be considered for revision because it is prescriptive, based on currently operating reactors, and may not be needed for safety in all cases ( e.g., for new advanced reactor designs). The NRC is proposing to change § 55.31(a)(5) to add a provision that in lieu of five significant control manipulations, the facility licensee may determine the minimum number of required control manipulations in accordance with its Commission-approved training program. This change would improve flexibility in those situations and ensure that an appropriate alternative is established.

B. Alternative Examination Standards

The NRC would revise paragraph (a) and delete paragraph (d) of § 55.40, “Implementation,” to allow the use of criteria other than the criteria in NUREG-1021, “Operator Licensing Examination Standards for Power Reactors,” and NUREG-1478, “Operator Licensing Examiner Standards for Research and Test Reactors,” to avoid the NRC having to revise the NUREGs for each new plant design and/or licensees having to submit exemptions that seek to use a document other than these NUREGs. Requiring the use of these NUREGs has been the means of complying with the AEA, which requires the Commission to “prescribe uniform conditions for licensing individuals as operators” for different classes of facilities, but the statutory requirement can still be achieved without having the examination standards for every plant design contained in a NUREG. For example, facility licensees could submit criteria for NRC review in an examination standards topical report.

C. Written Examinations and Operating Tests

The NRC would revise §§ 55.41(b), 55.43(b), and 55.45(a) to allow greater flexibility and clarity for the content of written examinations and operating tests. Currently, these regulations require that written examinations and operating tests include a representative sample of the topics listed therein, to the extent applicable to the design of the facility. These topics may not include all of the safety-significant topics applicable to plant designs that are not water-cooled reactors, so there may be other safety-significant topics that should be sampled on written examinations and operating tests for ( printed page 60716) operator license applicants at non-LWR plants but are not currently covered by these regulations. This could create ambiguity and confusion, resulting in exemption requests. The proposed edits to §§ 55.41(b), 55.43(b), and 55.45(a) would address these concerns.

D. Simplification of Simulator Requirements

The NRC would remove the terms “plant-referenced simulator” and “Commission-approved simulator” from 10 CFR part 55. The term “simulation facility” would remain, and its definition in § 55.4, “Definitions,” would be revised. The NRC would not add any new requirements for simulation facilities; the performance-based requirements that a simulation facility must meet in order to be used for operating tests, requalification, or meeting experience requirements would remain in § 55.46, “Simulation facilities.” These changes would clarify for facility licensees of reactors under construction the requirements that a simulation facility must meet to be used for operating tests, requalification, or meeting experience requirements. As long as a simulation facility meets these proposed requirements, it could be used for those purposes; there would be no need to determine whether the simulation facility is a plant-referenced simulator or if it instead must be approved as a simulation facility by the Commission. The NRC would confirm that a simulation facility complies with regulatory requirements by inspecting simulator performance test results and other simulator data as necessary. Additionally, the NRC has observed an increased use by the industry of “glasstop” simulators in training. “Glasstop” simulators consist of one or more touch-sensitive panels that mimic the analog and digital controls, alarms, and displays found in the main control room of a nuclear power facility. They do not look like a full-scale control room, but they can replicate plant systems accurately. The proposed changes would clarify for facility licensees of operating reactors the requirements to use glasstop simulators for additional purposes, such as conducting portions of the operating tests required by § 55.45. As long as a glasstop simulator meets the proposed requirements, it could be used for those purposes; there would be no need to determine whether the glasstop simulator is a plant-referenced simulator or if it instead must be approved as a simulation facility by the Commission. The NRC would confirm that a simulation facility complies with regulatory requirements by inspecting simulator performance test results and other simulator data as necessary. Minor conforming changes would be made to § 53.725(c).

E. Requalification Examination and Program Periodicity Requirements

In § 55.59(a), “Requalification requirements,”, the NRC requires each licensed operator to successfully complete a requalification program developed by the facility licensee and to pass a comprehensive written examination and an annual operating test as part of the requalification program. Paragraph (h) of § 55.53, “Conditions of license,” also contains a condition applicable to all operators' licenses that the licensee shall complete the requalification program, which includes passing the annual operating tests and biennial written examination. This is redundant to § 55.59(a).

Section 55.59(a) also lists the topics that must be included on the written examination and operating test; however, this requirement is a program requirement, and it would be more appropriate to include under the program evaluation requirements in § 55.59(c)(4), “Evaluation.”

Additionally, in § 55.59(c), “Requalification program requirements,” the NRC requires that the requalification program be conducted for a continuous period not to exceed 2 years, and that each 2-year program ( i.e., cycle) be followed by a subsequent 2-year program. The NRC proposes to allow a facility licensee, using the systems approach to training process, to appropriately determine a different length of time for its requalification program and different requalification examination periodicities.

The NRC identified the foregoing regulations as ones that should be considered for revision because they are prescriptive and may not be needed for safety in all cases. The proposed changes to § 55.59 would improve flexibility by allowing alternatives where they are justified by the facility licensee in its approved requalification program.

Conforming changes are proposed to 10 CFR part 53. Furthermore, the NRC proposes to amend § 53.730(g)(3) to align requalification program implementation requirements to those in § 50.54(i-1). Currently, § 53.730(g) states that requalification programs must be approved by the NRC as part of its approval of the OL or COL for the plant, and § 53.730(g)(3) requires that the requalification program be implemented upon commencing the administration of initial examinations. However, the administration of initial examinations could potentially begin before an OL is required to be approved. The proposed change would prevent this situation.

F. Requirements for Licensed Medical Examiners

The NRC would revise §§ 55.21, “Medical examination,” and 55.23, “Certification,” to allow either a physician or a licensed medical examiner to perform and certify the results of the medical examinations required for licensed operators and applicants for operator licenses. Currently, the medical examinations may only be certified by a physician, as defined in § 55.4, meaning an individual licensed to dispense drugs in the practice of medicine.

The NRC identified these requirements as anticompetitive and overly restrictive. The NRC proposes to allow a licensed medical examiner ( i.e., a medical professional, other than a physician, licensed by their state to independently perform medical examinations for the purpose of determining an individual's medical health and general condition) to perform and certify the results of the medical examinations required for licensed operators and applicants for operator licenses. This proposed change would reduce burden by expanding the pool of available medical professionals, especially in rural areas where doctors already may be in relatively shorter supply. If more nuclear power plants are constructed, more operators will be needed, and they will need medical examinations. Allowing medical professionals licensed by their state to perform medical examinations to do the operator examinations would likely help alleviate demand for doctors, especially in areas where they may be in short supply.

Conforming changes would be implemented in §§ 53.765 and 53.775(b)(2).

G. Changes to Treatment of Medical License Conditions

The NRC would revise §§ 55.25, “Incapacitation because of disability or illness,” 55.27, “Documentation,” 55.33, “Disposition of an initial application,” and 55.53 to allow for certain medical conditions to be covered under a new, generic condition of licenses in § 55.53. Currently, applicants for a license must have a medical examination that must be certified by a physician on NRC Form 396, “Certification of Medical Examination by Facility Licensee.” The certifying physician must determine that the applicant either meets all health ( printed page 60717) requirements or, if the applicant does not meet all health requirements, that any medical conditions can be accommodated through specific license conditions ( e.g., a license condition that says something akin to, “you shall wear corrective lenses while performing licensed operator duties,” for applicants who do not meet the minimum vision requirements). If the applicant's Form 396 indicates that the applicant has no medical issues requiring accommodation via a license condition, then the applicant can be found by the Commission to meet the requirements of § 55.33(a)(1). If the applicant's Form 396 indicates that the applicant has one or more issues that can be accommodated by one or more license conditions, then the Commission may issue a license with the specific license conditions that are necessary to accommodate the medical condition(s) ( e.g., “you shall take medications as prescribed,” for an applicant who has diabetes that is controlled by medication). Once a license is issued, the license must be amended if either any new, permanent medical conditions arise that need to be accommodated by one or more license conditions, or if a medical condition changes such that one or more license conditions are no longer necessary ( e.g., the operator gets LASIK surgery and no longer needs corrective lenses, rendering a license condition to wear corrective lenses unnecessary).

The NRC uses a qualified medical expert to review certain medical documents for applicants and operators to ensure that the license conditions proposed by the examining physician are sufficient to accommodate the identified medical conditions. The process is as follows:

a. For new applications, the NRC reviews the applicant's NRC Form 396.

b. If corrective lenses, hearing aids, and/or use of a continuous positive airway pressure device are checked as necessary license conditions, then the NRC will not send the documentation forward for review by a qualified medical expert. The NRC will then issue a license with the corresponding specific license condition(s).

c. If any other license conditions are identified as being necessary, then the NRC sends the documentation to a qualified medical expert, who reviews the applicant's medical information ( i.e., the “medical evidence” submitted on NRC Form 396) to make sure that the proposed license condition(s) will accommodate the medical condition(s). The qualified medical expert documents the results of his or her review and provides it to the NRC. The NRC issues the license with the necessary, specific conditions listed on the license.

d. If a licensed operator develops a new, permanent medical condition that must be accommodated by one or more license conditions, then they must inform the NRC and submit a new NRC Form 396 to request any new, specific license conditions be added to the license. The NRC uses the same process described above for new applications for license amendments.

To streamline this process, the NRC proposes to add a new, generic condition of licenses to § 55.53 that would eliminate the need for specific license conditions for certain medical restrictions ( i.e., corrective lenses, hearing aids, and use of a continuous positive airway pressure machine). The proposed new generic condition in § 55.53 would state that a licensee must comply with all license conditions identified on NRC Form 396 that are designated as covered by § 55.53(m), and NRC Form 396 would identify the medical restrictions ( i.e., license conditions) that are covered by § 55.53(m). The number of licenses with specific conditions for corrective lenses and/or hearing aids that could be replaced with the proposed new generic license condition are as follows per region/office (note: license conditions to use a therapeutic device are not included in the count below since they may or may not be limited to continuous positive airway pressure machines, so these numbers are slightly lower than the numbers of specific license conditions that would actually be able to be replaced):

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

On average, about 39% of licenses with specific license conditions would be “covered” under the proposed new general license condition in § 55.53(m).

Furthermore, there has been ambiguity for many years regarding what is considered a “permanent” medical condition. The NRC receives correspondence from facility licensees regarding temporary medical conditions; these correspondence are not required. The NRC proposes to amend §§ 55.25 and 55.27 to specify how facility licensees should accommodate temporary medical conditions, as distinct from permanent medical conditions, and to clarify that these do not need to be treated in the same manner as permanent medical conditions.

H. Changes to Operator License Reporting Requirements

As discussed in § 50.74, “Notification of change in operator or senior operator status,” three reports are currently required for changes to operator status. NEI discussed these reports in a May 6, 2020, letter on the Retrospective Review of Administrative Requirements (RROAR) (ML20128J340) and recommended deleting them. The reporting requirements in § 50.74(a) and (b) address the situations where the 10 CFR part 55 license holder's employment at the facility has been terminated or is no longer needed. The NRC does not propose to remove these two reporting requirements. The facility licensee should be responsible for reporting the required information to the NRC because (1) under this proposed rule, operator licenses would no longer be issued with license terms, and (2) the NRC needs to know the status of the operator licenses that it issues (including who actually holds operator licenses and, thus, is required to meet the requalification requirements and medical examination requirements). However, the NRC proposes to improve clarity by including these reporting requirements in 10 CFR part 55 instead of in 10 CFR part 50. The third reporting requirement, in § 50.74(c), is proposed to be deleted because it is redundant to §§ 55.25, and 55.25 would be amended to remove reference to § 50.74 and replace it with § 55.5, “Communications.”

Additionally, § 55.53(g) requires a licensed operator to notify the Commission within 30 days of conviction of a felony, and § 55.59(b) requires a licensed operator to submit evidence of completion of additional training to the Commission when necessary. The NEI letter dated May 6, 2020, regarding RROAR included a proposed change to § 55.53(g) to remove the burden of notifying the NRC from the individual operator. The NRC agrees and proposes to amend these sections to allow licensed operators to provide information to the NRC via the facility licensee rather than requiring the operator to provide it to the NRC directly. Conforming changes for § 55.53(g) would be made to §§ 53.785(h) and 53.810(f). Conforming changes for § 55.59(b) are not needed in 10 CFR part 53. ( printed page 60718)

I. Revision to Training Requirements in § 50.120

Paragraph (b)(1) in § 50.120, “Training and qualification of nuclear power plant personnel,” incudes an 18-month timeframe, based on the date of fuel load, for the establishment, implementation, and maintenance of a training program that meets the requirements of § 50.120(b)(2) and (3). This timeframe is not necessary for safety, and since the requirement was first established, ambiguity has existed as to what exactly is required to meet the 18 months before fuel load criteria. Therefore, the NRC proposes to change § 50.120 to clarify that the requirement to establish, implement, and maintain the training program must be met with sufficient time to ensure that qualified personnel are available to do their assigned jobs. This proposal was also included in a letter from NEI titled, “Industry Recommendations on Accelerating NRC Reform,” dated July 31, 2025 (ML25212A197).

Additionally, § 50.120 prescribes that specific training programs be implemented, including for a shift technical advisor. In 2020, NuScale submitted a topical report, which was approved by the NRC on May 26, 2021 (ML21231A286), to support future exemptions from the requirement to have a shift technical advisor training program because NuScale plants would not have a shift technical advisor. The NRC proposes to add “as applicable” to the list of personnel to be covered by the training program so as to avoid the need for future exemptions in cases where certain categories of plant personnel are determined not to be necessary at a plant.

Section 53.830, “Training and qualification of commercial nuclear personnel,” is already similar to proposed § 50.120, with a minor difference: § 53.830 states that prior to initial fuel load (or, for a fueled manufactured reactor, prior to initiating the removal of the features to prevent criticality required under § 53.620(d)(1)), each holder of an OL or COL must, with sufficient time to provide trained and qualified personnel to operate the facility, establish, implement, and maintain a training program that demonstrates compliance with the requirements of § 53.830(c) and (d). It imposes two time-based requirements to have the training programs established: prior to fuel load/initiating removal of features to prevent criticality, and with sufficient time to provide qualified personnel to operate the facility. The NRC proposes to remove the former requirement because there is no reason to also require the training program to be implemented prior to fuel load, as this activity would require sufficiently trained and qualified or even licensed plant personnel, and that would be covered by the requirement to have trained and qualified personnel in time to perform jobs that require trained and qualified personnel.

J. Elimination of Wait Times for Re-Applications in § 55.35

Paragraph (a) of § 55.35, “Re-applications,” requires an applicant for an operator license whose application has been denied because of failure to pass the written examination, operating test, or both, to wait 2 months after the first denial, 6 months after a second denial, and then 2 years for each subsequent denial. These wait times are unnecessary and create additional burden for operator license applicants and facility licensees. Therefore, the NRC proposes to amend this requirement to allow applicants to file re-applications regardless of the time that has elapsed since the previous application denial.

K. Operating Tests

In § 55.40, the NRC currently provides the option for power reactor facility licensees to prepare, proctor, and grade the written examinations required by §§ 55.41, “Written examination: Operators,” and 55.43, “Written examination: Senior operators,” and to prepare the operating tests required by § 55.45, “Operating tests,” subject to conditions specified therein. Section 55.40 also says that the NRC shall prepare, proctor, and grade examinations and tests for non-power reactor facility licensees. These regulations became effective on October 20, 1999, in the final rule, “Initial Licensed Operator Examination” (64 FR 19868). The Commission did not then permit licensees to administer and grade operating tests (with NRC examiners observing facility employees) in part due to a concern with the reduced NRC involvement and the potential for impact on independence, objectivity, and public perception. In SECY-96-206, “Rulemaking Plan for Amendments to 10 CFR part 55 to Change Licensed Operator Examination Requirements” (ML20134C900), the NRC staff discussed the option of having NRC examiners observe facility employees conduct the simulator scenarios and a sample of the walk-through tasks for the operating tests for each applicant: “[This option] would have had the least NRC involvement and, therefore, the greatest potential impact on independence and public perception. By not directly observing each applicant, the NRC examiner would be in a weaker position to judge the performance of the applicant so as to determine which applicants should pass or fail if a crew responded incorrectly during a simulator scenario.” In SRM-SECY-96-206, dated December 17, 1996 (ML20132C429), the Commission directed the NRC staff to continue to administer and grade the operating tests.

Today, nearly all power reactor operators' license written examinations and operating tests are prepared by facility licensee staff, and power reactor operators' license written examinations are frequently proctored and graded by facility licensee staff. As required by current regulations, operating tests are administered by an NRC examination team with the support of facility licensee personnel ( e.g., facility licensee personnel run the simulator used for power reactor simulator scenarios, taking direction from the NRC chief examiner on when the simulator should initiate the events per the pre-approved simulator examination guide), and the NRC examiners conduct the grading of their designated applicants, with no input from facility licensee personnel.

Certain power and non-power reactor stakeholders have recently indicated an interest in also proctoring and grading the operating tests for their operator license applicants, such that the NRC's role would shift from administering the operating tests and grading applicants to overseeing the facility licensee staff in its administration and grading of the operating tests. In addition, in its July 31, 2025, letter to the NRC, NEI recommended this change.

Moreover, industry now possess substantial experience in administering, proctoring, and grading examinations as part of both the Commission-approved INPO-accredited training process—which includes audit examinations that closely model NRC initial exams—and the long-standing NRC-mandated requalification program, which requires licensees to administer and grade annual operating tests and biennial written examinations subject to NRC inspection and oversight. Although the industry has not prepared the operating test grading documentation for the NRC initial operator licensing examination, the staff has developed detailed guidance in NUREG-1021 to support licensee preparation and documentation of initial exam grading, and NRC examiners would continue to oversee and evaluate licensee performance, enabling prompt identification and correction of any gaps that may arise during implementation. ( printed page 60719)

Section 55.40 is currently silent regarding who will administer and grade the operating tests for power reactor facilities, but it is implied that only the NRC will do that given that the regulation specifies other activities that power reactor facilities may do (and it is explicitly stated that way in NUREG-1021, revision 12). Therefore, this proposed rule would amend § 55.40 to specifically allow facility licensees to administer and grade operating tests subject to certain conditions.

L. Plant Walkthrough Requirements in § 55.45

In § 55.45(b), “Implementation—Administration,” the NRC requires that the operating test be administered in a plant walkthrough and in either a simulation facility that the Commission has approved for use after application has been made by the facility licensee under § 55.46(b), “Commission-approved simulation facilities and Commission approval of use of the plant in the administration of the operating test,” a plant-reference simulator, or the plant, if approved for use in the administration of the operating test by the Commission. Section 55.40(a) requires, in part, that the NRC use the criteria in NUREG-1021 to prepare the operating tests required by § 55.45, and to evaluate the operating tests prepared by power reactor facility licensees. As discussed in NUREG-1021, revision 12, section ES-3.1, “Overview of the Operating Test for Operator Licensing Initial Examinations,” the walkthrough portion of the operating test consists of a set of job performance measures. Each applicant for an operator's license must complete a certain number of job performance measures in the plant.

The plant walkthrough requirement was the subject of exemption requests for Vogtle Electric Generation Plant (VEGP) Unit 3 and for Virgil C. Summer Nuclear Station (VCSNS) Unit 2. In the June 30, 2016, Federal Register notice (FRN) of the NRC's granting approval of the exemption for VEGP Unit 3 (81 FR 42745) and in the August 22, 2016, FRN for the NRC's granting approval of the exemption for VCSNS Unit 2 (81 FR 56704), the NRC also approved alternatives to the in-plant methods of testing described in NUREG-1021. Specifically, the NRC approved the use of discussion and performance evaluation methods in combination with plant layout diagrams, maps, equipment diagrams, pictures, and mock-ups while the plant is under construction.

A substantial number of a plant's systems must be constructed before doing a plant walkthrough during the NRC initial licensing examination so that the plant walkthrough portion of the operating test is not predictable. Predictable examinations may prevent the examiner from distinguishing applicants who have mastered the required knowledge and skills from those who have not. However, the completion of plant construction occurs relatively close in time to the scheduled date for fuel loading and later operation of the facility. Thus, administration of the NRC initial licensing examinations would need to occur relatively close in time to the scheduled date for fuel loading at the plant. Delaying the administration of the NRC operating test, which includes in-plant job performance measures, until plant construction is complete is not desirable.

The NRC examination will likely overlap with preoperational testing activities; applicants may be required to participate in preoperational testing at the same time the NRC examination is administered. Insufficient NRC examination throughput ( i.e., the number of applicants that pass the NRC examination) could cause preventable delays in the facility licensee's ability to begin fuel loading. There might be missed opportunities for early identification and timely incorporation of lessons learned into the operator training and licensing process without a safe and deliberate approach to administering examinations to smaller groups of applicants in succession. The NRC's ability to administer examinations for the large number of applicants necessary to staff a new single large LWR unit (40-50 operators) in time for fuel load, while also carrying out examinations and inspections at operating reactor sites, could be challenged.

Therefore, the NRC proposes to amend the plant walkthrough requirements in § 55.45(b) to allow alternatives to the in-plant methods of testing while the plant is under construction.

M. Examination Waiver Requirements

Under § 55.47, “Waiver of examination and test requirements,” the NRC may waive any or all of the requirements for a written examination and operating test. As discussed in SECY-22-0052 and approved in SRM-SECY-22-0052, the NRC is proposing changes to the waiver criteria to allow waivers in the circumstances of adding additional units to a standard design. Additionally, § 53.780(f) would be revised with conforming changes.

N. Removing Obsolete Information From § 55.5

In § 55.5, instructions are provided for submitting information to the NRC. Currently, the electronic information exchange (EIE) is listed as an example of an electronic submission method. However, the NRC plans to transition away from the EIE and replace it with other options for the electronic submission of documents ( e.g., the NRC external portal). Therefore, the NRC proposes to remove the term “EIE” throughout 10 CFR part 55 as an example of a method of electronic submission.

O. Elimination of License Renewal for 10 CFR Part 55 Operators' Licenses

Historically, the requirement for renewal of operators' licenses afforded the NRC the opportunity to re-evaluate the licensing basis of licensed operators. The purpose of license renewal was to assess the likelihood of an operator's continued ability to safely operate the facility.

The NRC identified that eliminating license terms and, thus, the requirement for license renewal, would meet goals stated in E.O. 14300 and E.O. 14192, “Unleashing Prosperity Through Deregulation.” Specifically, if the number of nuclear plants and licensed operators were to increase, the amount of renewals would also increase, requiring an expenditure of more licensee and NRC resources. Removing the requirement for licensed operators to request renewal of their licenses every 6 years would be a reduction in burden for the licensed operators, facility licensees, and the NRC. Therefore, the NRC is proposing to eliminate license terms and the requirement for license renewal. Conforming changes would also be made to 10 CFR part 53.

The NRC has previously reconsidered license terms: in December 2024, the NRC eliminated license terms for certain NPUFs (see 89 FR 106234). The NRC determined that the overall objectives of license renewal can likely be achieved through existing oversight activities while eliminating the administrative burden associated with renewal. This proposed approach would be consistent with the NRC's goal of efficient and effective licensing. Additionally, this is not the first time that the NRC has amended renewal requirements for operators' licenses: license terms were extended from 2 years to 6 years in the final rule, “Operators' Licenses and Conforming Amendments,” issued on March 25, 1987 (52 FR 9453). ( printed page 60720)

P. Generally Licensed Reactor Operators for Self-Reliant Mitigation Facilities

Part 53, subpart F, describes a class of facility, the self-reliant mitigation facility (SRMF). For a facility to be classified as a SRMF, the NRC must find that certain criteria established in 10 CFR part 53 have been met. For SRMFs, specific licenses for operators ( i.e., individual operator and senior operator licenses) are not required. Instead, the facility licensee may rely on a general license for reactor operators issued by the NRC under 10 CFR part 53.

Part 53 is an optional framework, and 10 CFR parts 50 and 52 are also options for an NRC applicant. As of December 2025, three NRC applicants that are using or intend to use the current part 50 licensing process had proposed that their facilities are SRMFs, and, accordingly, expressed an interest in using the same provisions for a general license for reactor operators that are in 10 CFR part 53. Currently, there are no provisions for generally licensed reactor operators (GLROs) in the NRC's regulations other than in 10 CFR part 53. Accordingly, any applicant for an SRMF would need to request exemptions to use an approach comparable to a general license for reactor operators if it is not using 10 CFR part 53. Such an approach would be less efficient than adding to 10 CFR part 55 provisions equivalent to those in 10 CFR part 53. Therefore, the NRC proposes to add to 10 CFR part 55 optional provisions for GLROs, which would be applicable to 10 CFR parts 50 and 52 facility licensees of SRMFs.

Adding provisions for GLROs to 10 CFR part 55 would also require revising 10 CFR parts 50 and 52 to establish associated requirements related to staffing for GLROs akin to those requirements in 10 CFR part 53. The licensed operator staffing requirements in § 50.54(m) that apply to OL or COL holders do not account for GLROs and are not always necessary for safety.

Minor editorial conforming changes would be made to 10 CFR part 53.

10 CFR Part 75—IAEA Notifications

Paragraphs (b)(3) and (c) of § 75.13, “Safeguards on Nuclear Material—Implementation of Safeguards Agreements Between the United States and the International Atomic Energy Agency,” provide that a “request” for non-disclosure of information to the public made under § 2.390, “Public inspections, exemptions, requests for withholding,” not be treated as a “request” made under § 75.13 for confidential information not to be transmitted physically to the IAEA.

This rulemaking proposes eliminating § 75.13(c) because it is redundant with § 75.13(b)(3), as both contain identical language except that § 75.13(b)(3) references § 2.390 while § 75.13(c) references § 2.390(b).

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

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

IV. Discussion

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

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

A. 10 CFR Part 21, and §§ 50.55(e), and 53.605 Alignment

Section 50.55(e) has been modified over the last 30 years to become more aligned with 10 CFR part 21. Incorporating the two remaining differences from § 50.55(e) into 10 CFR part 21 would allow for the deletion of § 50.55(e). There are no identified disadvantages associated with combining these two regulations. The following proposed changes would remove redundant regulation and clarify regulatory requirements.

This proposed rule would delete § 21.2(b) and renumber the subsequent sections appropriately. This proposed rule would add § 21.2(a)(5) to address 10 CFR part 21 applicability to a construction permit, early site permit, and or manufacturing license and incorporate language from § 21.2(b).

In order to incorporate § 50.55(e) requirements into 10 CFR part 21, this proposed rule would add a new § 21.21(a)(3)(iii) with the following statement from § 50.55(e): Undergoes any significant breakdown in any portion of the quality assurance program conducted under the requirements of appendix B to 10 CFR part 50 which could have produced a defect in a basic component. These breakdowns in the quality assurance program are reportable whether or not the breakdown actually resulted in a defect in a design approved and released for construction, installation, or manufacture.

These proposed rule changes would add a new § 21.21(d)(1)(iii) and (d)(2) to include requirements for the holder of a facility construction permit subject to 10 CFR part 21, combined license, or manufacturing license, who obtains ( printed page 60721) information reasonably indicating that the quality assurance program has undergone any significant breakdown, to notify the Commission of the breakdown in the quality assurance program through a director or responsible officer or designated person.

This proposed rule would add a new § 21.51(b)(1)-(3). These paragraphs would state that the holder of a construction permit, combined license, and or manufacturing license must prepare and maintain records necessary to accomplish the purposes of § 21.51. Specifically, this requirement would include retaining records of evaluations of all deviations and failures to comply under § 21.21(a)(1) for the longest of 10 years from the date of the evaluation; 5 years from the date that an early site permit is referenced in an application for a combined license; or 5 years from the date of delivery of a manufactured reactor.

This proposed rule would add a new § 21.51(b)(4) for suppliers of basic components for holders of construction permits, combined licenses, and manufacturing licenses. This section would state that such suppliers of basic components must retain a record of all notifications sent to affected licensees or purchasers under § 21.21(c) for a minimum of 10 years following the date of the notification, and a record of the facilities or other purchasers to whom basic components or associated services were supplied for a minimum of 15 years from the delivery of the basic component or associated services.

This proposed rule would incorporate in 10 CFR part 21 applicable portions of the definition of “basic component” from 10 CFR part 50. Specifically, the language in paragraph (2) of the definition of “basic component” in § 50.2 would be added to the definition of “basic component” in § 21.3, “Definitions,” as a sixth item to include applicability to a construction permit holder, consistent with § 50.55(e).

This proposed rule would delete §§ 50.55(e) and 53.605 in their entirety. The proposed change of relocating specific §§ 50.55(e) and 53.605 requirements to 10 CFR part 21 would not impact licensees as compliance with the current §§ 50.55(e) and 53.605 would be sufficient to comply with the requirements proposed to be relocated. Additionally, there are currently no existing 10 CFR part 53 licensees.

B. Commercial Grade Dedication Definition for 10 CFR Parts 21, 52, and 53

This proposed rule would modify the definitions of “commercial grade item” and “dedication” in § 21.3 to include references to 10 CFR parts 52 and 53. This change would address (1) concerns raised after the 2007 10 CFR part 52 final rule, in which the NRC recognized that conforming changes to definitions in 10 CFR part 21 were not included to address 10 CFR part 52; and (2) the recent issuance of the 10 CFR part 53 final rule.

This proposed rule change for including references to 10 CFR parts 52 and 53 in the 10 CFR part 21 definitions of “commercial grade item” and “dedication” would not impact licensees and other stakeholders that are already implementing the commercial grade dedication acceptance process.

C. Executive Order 14267, Reducing Anti-Competitive Regulatory Barriers

This proposed rule would revise the § 21.3 definition of “dedicating entity” to address E.O. 14267. The proposed rule would also delete the word “third-party” with respect to who can perform the dedication, which has been determined to be anticompetitive, because it could limit competition among qualified suppliers.

The proposed rule would also clarify that a 10 CFR part 50, appendix B or appendix T compliant quality assurance program and 10 CFR part 21 program are needed when performing the commercial grade dedication process. In addition, the proposed rule would provide conforming changes to address the issuance of 10 CFR part 53 and the proposed rule to add 10 CFR part 50, appendix T. Revising the definition of “dedicating entity” would not impact licensees and other stakeholders that are already implementing the commercial grade dedication acceptance process. In addition, a clarification was added to the definition of “dedicating entity” to ensure that anti-competitive regulatory barriers are not introduced for fuel facilities. Fuel facilities, uranium enrichment and fuel fabrication facilities, plutonium processing and fuel fabrication plants, and spent fuel reprocessing facilities, that are licensed under 10 CFR part 70 and are required to develop an integrated safety analysis in accordance with subpart H to 10 CFR part 70, may have a dedicating entity perform dedication so long as the organization has a quality assurance program compliant with the licensee's approved configuration management program or management measure program, respectively. The proposed rule language is consistent with current licensee practices and aligns with the existing text in 10 CFR part 21 related to dedication for nonreactors.

D. Clarify “Basic Component” Definition

This proposed rule would modify the definition of “basic component” in § 21.3 to provide clarity to uranium enrichment, fuel fabrication, fuel conversion, plutonium processing and fuel fabrication plants or spent fuel processing facilities that are required, by regulation or license condition, to develop an integrated safety analysis in accordance with subpart H to 10 CFR part 70. For fuel facilities, this modification would focus the scope of commercial grade dedication to sole engineered items relied on for safety. Spent fuel reprocessing facilities and plutonium processing and fuel fabrication plants would also focus on sole engineered items relied on for safety and, additionally, include all engineered equipment that is relied on to prevent or mitigate high-consequences to individuals located beyond the controlled area boundary from a release of fission products and transuranics. While 10 CFR part 70 facilities are clearly within the scope of 10 CFR part 21, the NRC had previously received feedback that the definition of “basic component” in § 21.3, was difficult to interpret as it was written to have applicability to multiple other types of facilities and activities and did not offer sufficient specificity for such varied activities. In contrast, the power reactor definition of a basic component is specific to power reactor terminology. As a result, the fuel facility industry communicated through the years that they had difficulty in applying the definition as written, as made evident by numerous facility-specific definitions or exemptions. The revised definition of basic component would clarify the connection between subpart H to 10 CFR part 70, and 10 CFR part 21 using insights gained from years of operating experience, the current regulatory framework,^[4] and current industry procurement practices.

The NRC is proposing to take a risk-informed approach in the definition of basic component as applied to fuel facilities. The licensing basis for fuel facilities is well suited to taking such a risk-informed approach, as licensees that are subject to 10 CFR part 70, subpart H, are required to perform an integrated safety analysis. The integrated safety analysis designates items relied on for safety, consisting of passive engineered controls, active engineered controls, administrative controls, and enhanced administrative ( printed page 60722) controls (or a combination of these), to prevent or mitigate the consequences of events that could cause the performance requirements to be exceeded without the application of the items relied on for safety. These items relied on for safety are subject to requirements for configuration management, maintenance, training, reporting, and other quality assurance elements to ensure they are reliable and available to function as designed. In many cases, licensees have multiple items relied on for safety to independently prevent or mitigate a specific safety hazard. However, in some cases, licensees may have only one `sole' engineered item relied on for safety. In these cases, the NRC proposes that the sole item relied on for safety be designated a basic component.

By providing a new definition of basic components that would be specific to fuel facilities, plutonium processing and fuel fabrication plants or spent fuel reprocessing facilities, the NRC would provide clarity about the components that would be managed under a commercial grade dedication program and would need to be evaluated for deviations or defects. This definition, for fuel facilities, would focus dedication activities to components that are sole engineered items relied on for safety for preventing or mitigating intermediate-consequence events and high-consequence events including accidental criticalities. For spent fuel reprocessing facilities and plutonium processing and fuel fabrication plants, an additional criterion to this definition is proposed that would focus dedication activities on items relied on for safety that are necessary to prevent or mitigate hazards to members of the public located beyond the controlled area boundary unique to such facilities.

If a loss or degradation of any item relied on for safety results in failure to meet the performance requirements of § 70.61, the licensee is required to report the failure to the NRC. The new definition does not change the reporting requirements of either 10 CFR part 21 or 10 CFR part 70.

The definition of “basic component” would be modified to provide consistency with the definition of safety-related SSCs under 10 CFR part 53. Specifically, the changes to the definition of “basic component” made by the 10 CFR part 53 rule are inconsistent with the scope of what structures, systems, and components are considered safety-related in 10 CFR part 53. Therefore, this revision would correctly align the definition of “basic component” with the technology-inclusive framework of 10 CFR part 53.

Additionally, the definition of “basic component” would be modified to allow items designed and manufactured under a quality assurance program that is complying with the proposed appendix T to 10 CFR part 50 to be included in this definition. This change would not impact licensees as compliance with the proposed appendix T would be a voluntary alternative to appendix B to 10 CFR part 50.

E. Clarify That Commercial Grade Dedication Is Not Applicable to Nuclear Material Users

This proposed rule would remove the references to 10 CFR part 30 from the definitions of “commercial grade item” and “dedication” in § 21.3. This change would clarify that broad scope and specific licensees ( e.g., radiography, medical licensees, irradiators, and well logging) with licenses issued under 10 CFR parts 30, 31, 34, 35, 36, and 39, would not need to dedicate their basic components using the commercial grade dedication process. This proposed change would make the regulatory language consistent with industry practice. The NRC has not provided guidance to nuclear material users on the commercial grade dedication process, and enforcement of this portion of 10 CFR part 21 is of low safety significance. This proposed change would correct the regulation to ensure that nuclear material users continue to report defects on basic components but are not required to commercially dedicate them.

F. Remove 10 CFR Part 76 References

This proposed rule change would also remove references to 10 CFR part 76, which applied to gaseous diffusion plants. The proposed deletion of references to 10 CFR part 76 would not impact licensees and other stakeholders because the facilities subject to 10 CFR part 76 are no longer in operation and were transitioned to the regulatory oversight of the U.S. Department of Energy for decommissioning. The administrative removal of references to 10 CFR part 76 from 10 CFR part 21 would not impact NRC resources or external stakeholders. The NRC issued a direct final rule on December 3, 2025 (90 FR 55621), that amended various regulations to insert a conditional sunset date in response to E.O. 14270. This direct final rule added the conditional sunset date to 10 CFR part 76.

G. Update Notification Methods

This proposed rule change would update § 21.21(d)(3)(i) by replacing facsimile information with the NRC Headquarters Operations Officer (HOO) email address: hoo.hoc@nrc.gov. The proposed change would replace initial notification by facsimile with notification by email to the NRC HOO email address: hoo.hoc@nrc.gov as the preferred method, which would not impact licensees and other stakeholders as some are already implementing notification using the NRC HOO email.

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

A. Purpose and Scope

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

As part of its evaluation of each of the reporting requirements in § 50.72(b), the NRC considered the Principles of Good Regulation, the Be riskSMART framework, and the internal NRC evaluation and communications process for nonemergency event notifications (discussed in section 2.4 of the regulatory basis). The Principles of Good Regulation include independence, ( printed page 60723) openness, efficiency, clarity, and reliability and ensure that the agency focuses on its safety and security mission while appropriately balancing the interests of the NRC's stakeholders, including the public and licensees. The Be riskSMART framework provides a process to ensure the consistent application of risk information across different NRC program areas. Use of the Principles of Good Regulation and the Be riskSMART framework enabled the NRC to develop standard criteria by which to assess each of the reporting requirements. Under these criteria, the relative frequency of reporting of a specific event notification is not the sole factor in the justification for elimination of a requirement. The inclusion of the numbers of reactive inspections initiated due to event notifications provides some valuable risk insights ( i.e., unless deterministic criteria are identified, the decision to initiate a reactive inspection would be primarily based on conditional core damage probability). For some categories of reporting, the relative frequency of reports is low, challenging the basis described in PRM-50-116 about the administrative burden on licensees. Conversely, for some categories of reporting where there were a relatively greater number of reports, there were few of these events that tripped the risk threshold to initiate a reactive inspection, and most follow-up was conducted by resident inspectors under the baseline inspection program. Using the Be riskSMART framework, the data on frequency of event notification reports and the historical NRC response ( i.e., reactive inspection versus baseline inspection follow-up) was used to help inform the risk of eliminating these reports in the context of the original bases for the 4- and 8-hour reports. The NRC's application of these criteria is detailed in section 3.0 of the regulatory basis.

In this proposed rule, the NRC is not proposing changes to the following nonemergency reporting requirements:

a. Section 50.72(b)(1)—notification to the NRC within 1 hour of any deviation from the plant's technical specifications authorized by 10 CFR 50.54(x), if not already reported as part of a declared emergency;

b. Section 50.72(b)(2)(i)—notification to the NRC within 4 hours of the initiation of any nuclear plant shutdown required by the plant's technical specifications;

c. Section 50.72(b)(2)(iv)(B)—notification to the NRC within 4 hours of an event or condition that results in actuation of the reactor protection system when the reactor is critical, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation;

d. Section 50.72(b)(3)(ii)(A)—licensees to submit a report to the NRC within 8 hours of any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded;

e. Section 50.72(b)(3)(ii)(B)—licensees to submit a report to the NRC within 8 hours of any event or condition that results in the condition of the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety; and

f. Section 50.72(b)(3)(xiii)—notification to the NRC of an event that results in a major loss of offsite communications capability, offsite response capability, or emergency assessment capability.

Eliminating or delaying these reports would unacceptably degrade the NRC's ability to perform oversight and incident response functions for the potential significance or safety issue associated with those events. The NRC is also not proposing changes to the emergency reporting requirements found in appendix E, “Emergency Planning and Preparedness for Production and Utilization Facilities,” to 10 CFR part 50; the reporting requirements of § 50.73; or other event reporting requirements, such as those contained in 10 CFR part 30, “Rules of General Applicability to Domestic Licensing of Byproduct Material,” 10 CFR part 40, “Domestic Licensing of Source Material,” and 10 CFR part 70, “Domestic Licensing of Special Nuclear Material.”

B. Proposed Changes to § 50.72

The NRC proposes to amend the following nonemergency event reporting requirements in § 50.72. Eliminating these event reports would not be an unacceptable degradation of situational awareness because a delay in NRC awareness for the scope of these events and otherwise maintaining awareness via routine oversight processes would be commensurate with the safety significance of issues associated with these event reports. This proposed rule would eliminate the following reporting requirements.

a. Section 50.72(b)(2)(iv)(A)—the requirement to notify the NRC of an event that results or should have resulted in an ECCS discharge into the reactor coolant system as a result of a valid signal except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation;

b. Section 50.72(b)(2)(xi)—the requirement to notify the NRC of an event or situation related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made;

c. Section 50.72(b)(3)(iv)(A)—the requirement to notify the NRC of an event or condition that results in valid actuation of specified safety systems;

d. Section 50.72(b)(3)(v)—the requirement to notify the NRC of an event or condition that could have prevented fulfillment of the safety function of structures or systems needed to shut down and maintain the reactor in a safe condition, remove residual heat, control the release of radioactive material, or mitigate the consequences of an accident; and

e. Section 50.72(b)(3)(xii)—the requirement to notify the NRC of an event requiring the transport of a radioactively contaminated person to an offsite medical facility for treatment.

This proposed rule would also add new paragraph (a)(6) to § 50.72 to enable licensees to submit certain nonemergency event reports via methods other than the ENS.

i. Proposed Changes to § 50.72(b)(2)(iv)(A)

In § 50.72(b)(2)(iv)(A), the NRC currently requires that each nuclear power reactor licensee notify the NRC of an event that results or should have resulted in a discharge of the ECCS into the reactor coolant system resulting from a valid signal unless the actuation resulted from and was part of a pre-planned sequence during testing or reactor operation. The § 50.72(b)(2)(iv)(A) report is required to be made within 4 hours of occurrence of such an event.

The NRC proposes to eliminate this reporting requirement because these reports are infrequent, occurring 37 times across the entire nuclear power plant fleet between 2011 and 2021, and have not required prompt agency actions as discussed in section 14.2 of the regulatory basis. Agency records indicate that the NRC has never activated its response plan in response to these events absent an emergency declaration. Most instances where the underlying events result in a notification under this requirement will happen when the RPS is in service. ECCS actuations that also result in RPS actuation signals would still be captured by the RPS actuation notification requirement in the current § 50.72(b)(2)(iv)(B). Further, if the ( printed page 60724) underlying condition that resulted in the ECCS actuation meets the threshold for an emergency declaration as required by a licensee's emergency plan, such conditions would still be reportable within 1 hour under § 50.72(a)(3).

Finally, there may be some low-risk conditions where ECCS actuations occur that do not result in RPS actuations or meet licensees' pre-defined emergency declaration conditions. These low-risk actuations would still be captured in the LER reporting requirement in § 50.73(a)(2)(iv), which would continue to ensure openness by making information about these events publicly available. As those conditions are of low risk, they would not result in the need for prompt agency action, and reporting only via LER would be appropriate. Inspectors would likely identify and may follow up on these events during their normal oversight duties well before an LER was issued, depending on the safety significance of the event. Further discussion of the factors and options considered by the NRC can be found in section 14.0 of the regulatory basis.

ii. Proposed Changes to § 50.72(b)(2)(xi)

In § 50.72(b)(2)(xi), the NRC requires that licensees notify the NRC of any event or situation, related to the health and safety of the public or onsite personnel, or protection of the environment, for which a news release is planned or notification to other government agencies has been or will be made. Such an event may include an onsite fatality or inadvertent release of radioactively contaminated materials. The report is required to be made within 4 hours of occurrence of such event.

The NRC proposes to eliminate this reporting requirement because these events are generally not related to nuclear plant safety and, therefore, do not pose a threat to public health and safety. Agency records indicate that the NRC has never initiated an inspection or investigation in response to these events. Removing this requirement would not affect the NRC's ability to execute oversight, event response, or other regulatory functions.

The NRC and the public may first become aware of issues, such as fatalities or contamination, from news and social media. In addition, public or NRC situational awareness of significant items would continue to occur through the normal regional baseline inspections and subsequent inspection reports that already address the underlying condition resulting in offsite nonemergency event notifications. As an example, resident inspectors routinely sample and review condition reports generated on a daily basis under the corrective action program. These condition reports capture situations that would require reporting by the current criteria proposed to be eliminated by this rulemaking. The NRC would remain properly informed of these events and be able to communicate them to the public via quarterly baseline inspection reports, if appropriate. Responses to any public inquiries could be developed and coordinated through the NRC's Office of Public Affairs.

iii. Proposed Changes to § 50.72(b)(3)(iv)

In § 50.72(b)(3)(iv)(A), the NRC requires that nuclear power reactor licensees notify the NRC of an event or condition that results in valid actuation of any of the systems listed in § 50.72(b)(3)(iv)(B) except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. The report must be made within 8 hours of the occurrence of such an event. Section 50.72(b)(3)(iv)(B) lists the systems subjected to the reporting requirements in § 50.72(b)(iv)(A), and the NRC is proposing to delete this list as a conforming change. The systems listed in § 50.72(b)(3)(iv)(B) are the following: reactor protection system including reactor scram and reactor trip; general containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves; ECCS for pressurized water reactors including high-head, intermediate-head, and low-head injection systems and the low pressure injection function of residual (decay) heat removal systems; ECCS for boiling water reactors including high-pressure and low-pressure core spray systems, high-pressure coolant injection system, and low pressure injection function of the residual heat removal system; boiling water reactor core isolation cooling system, isolation condenser system, and feedwater coolant injection system; pressurized water reactor auxiliary or emergency feedwater system; containment heat removal and depressurization systems, including containment spray and fan cooler systems; and emergency ac electrical power systems, including emergency diesel generators (EDGs), hydroelectric facilities used in lieu of EDGs at the Oconee Station, and boiling water reactor dedicated Division 3 EDGs.

The NRC proposes to eliminate this reporting requirement because these reports have not required prompt agency oversight or enforcement action within a day as discussed in Section 19.2 of the regulatory basis. The routine follow-up of plant issues already performed by resident inspectors is sufficient to ensure that the NRC takes appropriate near-term actions in response to these events. Historically, the NRC has not needed to take prompt actions for event notifications reported under this criterion outside of routine event response, thus, retaining this criterion is not warranted. Formal notification through LERs under § 50.73(a)(2)(iv) would provide the NRC with more detailed information on the event and inform longer-term actions to respond to such events. Further discussion of the factors and options considered by the NRC can be found in section 19.0 of the regulatory basis described in the “Part 50—Reporting Requirements for Nonemergency Events for Nuclear Power Plants” section of this document.

iv. Proposed Changes to § 50.72(b)(3)(v)

In § 50.72(b)(3)(v), the NRC requires that nuclear power reactor licensees notify the NRC of an event or condition that, at the time of discovery, could have prevented the fulfillment of the safety function of structures or systems that are needed to shut down the reactor and maintain it in a safe shutdown condition, remove residual heat, control the release of radioactive material, or mitigate the consequences of an accident. The report must be made within 8 hours of the occurrence of such an event.

The NRC proposes to eliminate this reporting requirement because these reports have been infrequent and rarely require prompt agency oversight or enforcement action. The routine follow-up of plant issues already performed by resident inspectors is sufficient to ensure that the NRC takes appropriate near-term actions in response to these events. Formal notification through LERs under § 50.73(a)(2)(iv) would provide the NRC with more detailed information on the event and inform longer-term actions to respond to such events. Further discussion of the factors and options considered by the NRC can be found in section 20.0 of the regulatory basis.

v. Proposed Changes to § 50.72(b)(2)(xii)

In § 50.72(b)(3)(xii), the NRC requires that nuclear power reactor licensees notify the NRC of any event requiring the transport of a radioactively contaminated person to an offsite medical facility for treatment. This report is required to be made within 8 hours of the occurrence of such an event.

The NRC proposes to eliminate this reporting requirement because events ( printed page 60725) covered by § 50.72(b)(3)(xii) are generally not related to plant safety and licensees and offsite emergency responders have arrangements and agreements in place to safely transport contaminated individuals without compromising public health and safety. Specifically, the regulations in paragraph (b)(12) of § 50.47, “Emergency plans,” and appendix E to 10 CFR part 50 require licensees to have arrangements and agreements in place for transportation and medical services for radiologically contaminated individuals, which includes notification of the offsite medical facilities. The regulations in paragraph (a)(12) of 44 CFR 350.5, “Criteria for review and approval of State and local radiological emergency plans and preparedness,” and the Federal Emergency Management Agency's (FEMA) Radiological Emergency Preparedness Program Manual provide policies and guidance to offsite response organizations, including to have these arrangements with licensees as well. The NRC regulations in § 50.47(b)(14) and appendix E to 10 CFR part 50 and the FEMA regulations in 44 CFR 350.5(a)(14) also contain training provisions for licensees and offsite emergency responders, respectively. Therefore, eliminating the reporting requirement in § 50.72(b)(3)(xii) would not pose a threat to public health and safety.

Agency records indicate that the NRC has never initiated an inspection or investigation in response to these events and would still be able to successfully execute its oversight, event response, and regulatory functions. Further discussion of the factors and options considered by the NRC can be found in section 21.0 of the regulatory basis.

vi. Proposed Changes to § 50.72(a)

The NRC is proposing to allow licensees to use alternative methods to notify the NRC of nonemergency events listed in § 50.72(b)(2)-(3). Currently, each operating nuclear power reactor licensee is required, in part, under § 50.72(a)(1) to notify the NRC Headquarters Operations Center (HOC) through the ENS of the declaration of any of the emergency classes specified in the licensee's approved emergency plan and nonemergency events specified in § 50.72(b). The ENS lines are direct access telephone lines that bypass switching in the telephone service's local central office. These lines are installed in the licensee control rooms as well as their technical support center and emergency operations facility. If the ENS is inoperative, § 50.72(a)(2) requires the licensee to make the required emergency notifications through a commercial telephone service, another dedicated telephone system, or any other method enabling a report as soon as practical to the HOC at the telephone numbers specified in appendix A to 10 CFR part 73, “Physical Protection of Plants and Materials.” Although use of NRC Form 361, “Reactor Plant Event Notification Worksheet,” is voluntary, it provides a standard structure for licensees to communicate to the HOC the information required by § 50.72.

In the early 1980s, when the NRC first required and implemented the ENS, telephones were the only widely available means of communication available between the NRC and its licensees. Since then, several other means of communication have become widely available, including facsimile, email, text messaging, instant messaging, and online portals. Licensees have used facsimile or other electronic means to submit a completed version of Form 361 to the NRC with specific event information as a supplement to their telephone reporting required by § 50.72(a)(1). Often, the completed version of NRC Form 361 would be sufficient to notify the NRC of the event. However, licensees must continue to use the ENS to call the HOC because of the requirement to do so in § 50.72(a)(1). In cases for which NRC Form 361 would be sufficient, the mandatory ENS phone call would be unnecessary for many nonemergency reports.

The NRC is proposing to add a new § 50.72(a)(6) to allow licensees to submit nonemergency event reports required under § 50.72(b)(2)-(3) using other methods as an alternative to using the ENS, provided the licensee submits the report to the NRC HOC within the required timeframe and confirms the NRC's receipt of the report. For example, the NRC offers certain persons an online method of submitting reports to the NRC, although licensees are still required by regulation to communicate all events directly to the HOC through the ENS. Beginning in 2022, the NRC began allowing properly vetted licensee personnel to provide NRC Form 361 through an online portal. Upon submittal, the information is sent to the HOC for the Headquarters Operations Officers for follow up with the licensee, if needed, and process to the NRC's operating experience team (nuclear power plants only) for data retention, tracking, and trending. Under this proposed rule, licensees using this online portal for § 50.72(b)(2) or (3) nonemergency event reports would need only to confirm NRC receipt of the report and would not need to use the ENS.

The NRC would not allow licensees to submit nonemergency event reports required under § 50.72(b)(1) using methods permitted by proposed § 50.72(a)(6). The event listed in § 50.72(b)(1)—deviation from the plant's technical specifications, if not already reported as part of a declared emergency—is required to be reported within 1 hour of its occurrence because the event is ongoing, and actions are being taken to address immediate safety or security concerns. The nature of these reports is such that prompt actions or follow-up by the NRC may be warranted commensurate with safety significance. The HOC is the primary center of communication and coordination among the NRC, its licensees, and other stakeholders for operating events and needs to be made aware of these events within 1 hour. Event notifications required by § 50.72(b)(1) must be made through the ENS so the HOC promptly receives them.

C. NRC Situational Awareness and Public Openness

For events associated with reporting criteria that are proposed for elimination, NRC situational awareness would continue under the normal inspection activities as part of the Reactor Oversight Process (ROP). These events would be treated as other plant conditions of similar safety significance are treated and would be subject to each licensee's informal communication processes and the NRC's baseline sampling process that resident inspectors apply to all reported plant conditions. Resident inspection staff would continue to perform baseline inspections of selected conditions and document findings in inspection reports regardless of reportability under the nonemergency event notification criteria. As an example, resident inspectors routinely review daily plant condition reports under the corrective action program and through their routine plant status activities. Those baseline inspection activities capture conditions of potential public health and safety concern occurring at licensee sites, including situations that would be reported to offsite organizations. The proposed changes are not intended to increase the workload of resident inspectors or require resident inspectors to gather, verify, or communicate information beyond what is done through existing inspection activities.

Many ROP-related activities include actions to inform the NRC of items of interest within a sufficient timeframe, such as those specified in inspection ( printed page 60726) procedures, routine regional status meetings, and frequent informal communications between inspectors and regional management. For significant items discovered during these activities, regional staff perform reactive inspection evaluations in accordance with MD 8.3, “NRC Incident Investigation Program.” The MD 8.3 process is the NRC's primary internal tool for ensuring that significant operational events are investigated in a timely, objective, systematic, and technically sound manner. These activities, which occur independently of nonemergency event notifications, ensure that the NRC would remain informed, act in a timely manner, and communicate items of interest internally.

Although the proposed approach would eliminate prompt notifications for some events that have no or very low safety or regulatory significance, the NRC would maintain sufficient NRC situational awareness of these events, as appropriate, via regional baseline and emergency preparedness inspection activities, including available performance indicators. Notification of individual events to NRC Headquarters technical points of contact and the public may be delayed or not occur for a subset of issues. For example, a situation that results in an ECCS system actuation but not an associated RPS actuation is currently reportable under either a 4-hour or 8-hour report depending on the circumstances. Such a report is shared internally with NRC Headquarters technical points of contact and management within 24 hours of receipt for situational awareness. As proposed in this rulemaking, this event might not be formally reported to the NRC until a required LER is made 60 days after the event or not at all. However, regional inspection staff and management would become aware of these ECCS events within several days through performing routine baseline inspection activities. Other NRC technical points of contact and management may not be promptly notified for purposes of situational awareness but may become aware through other routine communication processes. As another example, critical emergency preparedness assessment capabilities are addressed in licensee EAL schemes, and a major loss of capabilities would result in a 1-hour emergency event notification per § 50.72(a)(3), thereby providing the NRC with sufficient situational awareness.

Openness and public awareness would be maintained under the normal regional baseline inspections and subsequent inspection reports that already address situations resulting in offsite notifications. The release of public information on some plant events may be delayed, or not occur, for some events that have low safety or regulatory significance. As an example, resident inspectors routinely review daily plant condition reports generated under the corrective action program. This review includes the underlying conditions that are currently required to be reported to the NRC but would not be immediately reported under this proposed rule, thereby ensuring the NRC would remain informed and able to communicate issues of safety or regulatory significance to the public via quarterly baseline inspection reports.

If a nonemergency report is not generated and interest from external stakeholders exists, then the NRC would consider communication of event information to stakeholders, where appropriate, following the current processes for communicating items of interest. For issues of public interest, responses could be developed and coordinated through the NRC's Office of Public Affairs. Internal interest could be managed using existing processes and regular operating experience briefs.

In summary, the inspection program would ensure sufficient openness and sharing of information with the public for significant issues subject to LER reporting criteria and, for items where findings are identified, through inspection reports. However, this information sharing would take longer than for the current event notification criteria and may not occur for items that have no associated LER reporting criteria ( e.g., major loss of communications) and for which regulatory findings are not identified.

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

New and advanced reactor technologies in a variety of sizes, geographic locations, technical complexities, and licensed footprints may be licensed under the long-standing frameworks in 10 CFR parts 50 and 52 or the recently published 10 CFR part 53. As the NRC prepares to license reactors that differ from the traditional, large light-water reactor designs that the United States has seen for decades, the NRC intends to establish and maintain regulatory frameworks that are risk-informed and technology-inclusive. Given the number of options regarding regulatory frameworks that an applicant may pursue to license a specific reactor technology, the NRC is proposing to apply regulatory enhancements to decommissioning licensing requirements across all three frameworks ( i.e.,10 CFR parts 50, 52, and 53). These enhancements are described in the following subsections.

A. Expansion of Partial Site Release Requests

The NRC proposes to add new § 52.111, “Release of part of a power reactor facility or site for unrestricted use,” as a specific provision for release of a part of a site for unrestricted use before license termination for part 52 licensees. The proposed § 52.111 would be nearly identical to the text in § 50.83, with specific applicability to holders of an operating license or combined license under 10 CFR part 52. The rule language from the existing § 50.83 and the regulations referenced within it are otherwise already applicable for 10 CFR part 52 facilities and no other changes to the language from the existing § 50.83 would be needed. The proposed changes are anticipated to provide 10 CFR part 52 licensees with a clearer and more efficient path towards releasing portions of their site prior to license termination and will align the regulations in 10 CFR part 52 with those in 10 CFR parts 50 and 53.

Minor conforming changes are also proposed for § 20.1401, “General provisions and scope,” in paragraphs (a) and (c). The proposed change to § 20.1401(a) would add the release of a part of a facility or site for unrestricted use in accordance with proposed § 52.111 to the list of facilities that the criteria in part 20, subpart E, “Radiological Criteria for License Termination,” apply to. The proposed change to § 20.1401(c) would limit the circumstances in which the Commission could require additional cleanup after release of part of a facility or site for unrestricted use in accordance with proposed § 52.111 to those instances where based on new information it is determined that the criteria of part 20 subpart E were not met and residual radioactivity remaining at the site could result in significant threat to public health and safety.

B. Reducing Regulatory Requirements for Prompt Decommissioning of Power Reactors

The NRC is proposing to expand the options by which nuclear power reactors may terminate their license. Because of the range of commercial and industrial utilization and production facilities that may become licensed under 10 CFR part 50, 10 CFR part 52, or 10 CFR part 53, the NRC is proposing to enhance the decommissioning licensing process to accommodate the unique aspects of a variety of facilities. ( printed page 60727) Currently, a nuclear power reactor licensee must submit a PSDAR within 2 years of permanent shutdown. The licensee may perform dismantlement and remediation activities permitted by § 50.59 or § 53.1540, “Updating licensing-basis information and determining the need for NRC approval.” Then, at least 2 years before expected license termination, the licensee must submit an LTP for NRC review and approval. The current decommissioning framework is predicated on the assumption that power reactors will take a significant amount of time, potentially up to 60 years, to terminate their license. However, the proposed regulations would more readily and efficiently accommodate prompt decommissioning of power reactors.

The NRC is proposing to amend the decommissioning licensing process to allow licensees to submit the LTP within 2 years of permanent cessation of operations (referred to as the “upfront LTP”). In this case, the licensee would not be required to submit the PSDAR. The goal of this proposed change is to provide an option for increased regulatory certainty, in the form of a license application and approval, at the beginning of the decommissioning process for reactors that choose to submit an upfront LTP. The NRC has received stakeholder feedback from the power reactor industry that licensees perform the majority of decommissioning activities “at risk,” without NRC review and approval, before the submission of the LTP. At the time of NRC review and approval of the licensing action (within 2 years of expected license termination), the licensee has already removed the reactor pressure vessel, cut and dismantled various reactor systems, and remediated building surfaces and land areas. For many facilities, these major decommissioning activities could potentially increase the amount and variability of residual radioactivity that must be remediated and surveyed before license termination. Consequently, this timing challenge can reduce the efficiency of the decommissioning licensing process, as licensees may need to implement additional corrective measures to meet regulatory requirements after certain decommissioning activities have been completed. For example, consider a hypothetical site has completed its final status survey of an excavation and subsequently backfilled it with offsite surface soil. If the NRC determines that the final status survey is incomplete or inaccurate, the licensee may not be able to selectively remove the clean backfill in order to redo the survey. In another example, a hypothetical site may perform radiation surveys of a building before demolition. If the NRC identified a concern with the survey results, the licensee may need to perform additional remediation activities near the building area, which may have already been surveyed, as well. In such cases, NRC-identified concerns of decommissioning activities performed “at risk” may delay license termination for certain facilities that had conducted such decommissioning activities before meeting all regulatory requirements.

Licensees that submit an upfront LTP within 2 years of permanent cessation of operations would benefit from increased regulatory certainty during the decommissioning process. Additionally, licensees opting for this approach would not be required to submit a PSDAR, as its three required components under § 50.82(a)(4)(i) ( e.g., schedule and description of the planned decommissioning activities, discussion of environmental impacts, and site-specific decommissioning cost estimate) are already captured within the eight elements of the LTP under § 50.82(a)(9)(ii), thereby streamlining the licensing process and avoiding duplication of effort. When the PSDAR and LTP were established in 1996 (61 FR 39301), the NRC noted that the PSDAR and LTP could be combined into one submittal, which would have to be submitted to NRC within 2 years of permanent shutdown. In this case, the requirements for both the PSDAR and LTP would apply to the combined document, including the requisite waiting period, public meeting, and approval by amendment of the LTP. However, no power reactor licensee, to date, has submitted a combined PSDAR and LTP. This proposed rule would streamline the licensing process for facilities that are able to prepare an LTP within 2 years of permanent cessation of operations.

This proposal is similar to the long-standing, proven decommissioning licensing process of submitting a Decommissioning Plan (DP) soon after permanent cessation of operations. This approach was the standard decommissioning licensing framework for power reactor licensees until the introduction of the PSDAR and LTP. The DP approach, which is currently the required licensing mechanism for nonpower production or utilization facilities and all nuclear materials facilities, requires licensees to submit the DP for NRC review and approval promptly after permanent cessation of operations. Under this approach, the licensee could not perform dismantlement activities until the approved DP amended the license and the NRC issued an associated order permitting decommissioning activities in accordance with the approved plan. However, the proposed revisions here differ from the DP in that power reactor licensees would be able to perform decommissioning activities consistent with § 50.59 or § 53.1540, as applicable, prior to the approval of the upfront LTP.

The NRC anticipates that future reactors licensed under 10 CFR parts 50, 52, and 53 may be smaller, less technically complex, and may operate for a shorter period of time than the current fleet of traditional reactors, while maintaining adequate records and decommissioning financial assurance. These facilities may plan for and begin decommissioning activities promptly after permanent cessation of operations, during operations, or even during application for an operating license. For example, consider a hypothetical SMR, licensed under 10 CFR part 52, which operates for 5 years in a remote location to provide emergency power to a local community after a natural disaster. Because the facility has a short operating history, has not undergone reactor refueling, and has a small license footprint, the licensee wishes to decommission promptly after cessation of operations. The licensee plans to terminate the license within 2 years. Under the current regulations, the licensee would either need to request an exemption from the requirement to submit a PSDAR or submit both the PSDAR and LTP after permanent shutdown. However, there would be no increased safety benefit from submitting both a PSDAR and LTP within this short timeframe. Under the proposed regulations, the licensee could focus on developing and submitting a quality LTP to promptly terminate the license in a safe manner. Accordingly, this proposed licensing enhancement would significantly reduce administrative burden on licensees who submit an upfront LTP.

The NRC foresees challenges for traditional large power reactor licensees in developing an LTP within 2 years of permanent cessation of operations. The long-standing PSDAR and LTP approach was developed specifically for large power reactors to transition to decommissioning with less regulatory burden and begin decommissioning activities with sufficient funding promptly. However, upon entering decommissioning, licensees must plan and safely perform large-scale dismantlement and remediation activities that have not been attempted ( printed page 60728) during operations ( e.g. removal of the reactor pressure vessel, dismantlement of the containment building). Traditional large power reactor licensees may need time to identify remaining dismantlement activities and develop the historical site assessment, sufficient characterization reports, and detailed plans for final radiation survey to be included in the LTP. The NRC does not anticipate that the current fleet of reactors would benefit from using the upfront LTP approach because the licensee would need to develop a large volume of technical information soon after operating a large facility for many decades. Rather, the upfront LTP would be most beneficial for licensees who expect to begin decommissioning promptly after permanent shutdown; have less complex dismantlement, decommissioning planning, and radiation survey activities; and can feasibly terminate the license in a foreseeable timeframe. Nevertheless, the NRC is proposing to provide the option to submit an upfront LTP to all power reactor facilities licensed under 10 CFR parts 50, 52, and 53. Because both decommissioning pathways are proven methods to meet NRC requirements for license termination, the agency does not propose to limit the upfront LTP option to a subset of licensees, but rather allow all power reactor licensees to determine which approach best suits their operational and planning needs.

Accordingly, this proposed rule has conforming changes to §§ 50.82(a)(5), 50.82(a)(7), 52.110(e), 52.110(g), 53.1070(e), and 53.1070(g), which reference submittal of the PSDAR. This proposed rule would update the requirements to reference the submittal required before or within 2 years of permanent cessation of operations. The NRC would retain the core content of the requirement, regardless of whether the licensee chooses to submit the PSDAR or LTP at that time.

The NRC is proposing to amend the duplicative requirement in §§ 50.82(a)(8)(iii) and 52.110(h)(3) related to submission of a site-specific decommissioning cost estimate. Both the PSDAR (in §§ 50.82(a)(4)(i) and 52.110(d)(1)) and LTP (in §§ 50.82(a)(9)(ii)(F) and 52.110(i)(vi)) require licensees to submit an updated site-specific decommissioning cost estimate. There is no situation in which a licensee will not have already submitted a site-specific decommissioning cost estimate within 2 years following permanent cessation of operations. Therefore, this regulation is redundant.

Currently, the header of paragraphs (a) and (b) of § 50.82, “Termination of license,” denote the applicability of the subsequent requirements to power reactor licensees and non-power production or utilization facility licensees, respectively. The NRC is proposing to move these indicators into the introductory paragraph of § 50.82, effectively restructuring the section to ensure cohesiveness and avoid confusion. The NRC will likely license a variety of power reactors under 10 CFR part 50 or other applicable licensing frameworks. Each of these licensing frameworks have their own set of decommissioning requirements. Therefore, it may be confusing for future power reactor licensees under a different part to determine if § 50.82(a) applies. The proposed introductory paragraph would note that the requirements in paragraph (a) apply to power reactor facilities licensed under 10 CFR part 50 and that the requirements in paragraph (b) apply to non-power production or utilization facility licensees under 10 CFR part 50. This proposed change would increase regulatory clarity for facilities licensed under 10 CFR part 50 as well as avoid confusion for power reactors licensed under other licensing frameworks.

C. Public Engagement Related to Decommissioning Licensing Actions

The NRC considers public involvement in, and information about, NRC activities to be a cornerstone of strong regulation of the nuclear industry. The NRC recognizes the public's interest in the regulation of nuclear activities and provides opportunities for the public to be heard. The NRC has an established record of engaging in open communications with members of the public, local community engagement groups, industry stakeholders, and tribal and foreign governments to facilitate public participation and hear feedback. At present, 10 CFR part 50 requires the NRC to hold public meetings in three instances related to the decommissioning process ( i.e., during review of a PSDAR, LTP, and partial site release request).

Following receipt of the PSDAR, the NRC currently publishes the PSDAR in the Federal Register and in a local forum such as a local newspaper, responds to all public comments received, holds a public meeting in the vicinity of the site, and publishes an associated meeting summary. These actions impose significant costs on the NRC and licensees, without commensurate safety benefit. The NRC does not approve the PSDAR but rather reviews the document for accuracy and completeness. The NRC uses the PSDAR to plan its decommissioning oversight activities, ensure compliance with decommissioning financial assurance requirements, and understand the environmental impacts of planned decommissioning activities. Historically, public comments submitted through the Federal Register and raised at public meetings related to the PSDAR have been related to issues other than the decommissioning of the specific facility. Public comments have rarely resulted in changes to a licensee's PSDAR. Therefore, based on decades of experience, this proposed rule would eliminate the requirement for the NRC to publish the PSDAR in the Federal Register and in a local forum such as a local newspaper, and solicit public comments on it. The NRC would retain the requirement for a public meeting upon receipt of the PSDAR to continue to engage the public. Upon receipt of the LTP, current regulations require the NRC to schedule a public meeting and publish a notice in the Federal Register and in a forum, such as local newspapers, which is readily accessible to individuals in the vicinity of the site. Regardless of the timing in which the LTP is submitted ( e.g., upfront LTP within 2 years of permanent cessation of operations or traditional LTP within 2 years of expected license termination), current regulations require, and the proposed regulation would continue to require, the NRC to engage with the public in this manner.

There are a variety of ways in which the public can engage the NRC to obtain access to information and provide written comments. Members of the public can view and follow NRC social media pages, register for NRC email updates, and view Really Simple Syndication (RSS) feeds to stay abreast of current news and information from the NRC. The NRC public website contains up-to-date information about each of the facilities undergoing decommissioning, including the name of the NRC project manager for the site. Members of the public have access to the public search function of the agency's official recordkeeping system, ADAMS, and may ask questions, send comments, or make recommendations about ADAMS by contacting the librarians in the PDR. A member of the public may submit a written request for copies of records that are not routinely made available as part of the agency's public record. Moreover, members of the public are able to obtain access to Federal agency records in accordance with the Freedom of Information Act and the Privacy Act.

Additionally, the NRC attends a variety of periodic meetings held by ( printed page 60729) local communities, state officials, and tribal governments. For example, the NRC routinely attends Nuclear Decommissioning Citizens Advisory Panel (NDCAP) meetings near facilities undergoing decommissioning. For a decommissioning action related to a specific facility, the NRC may hold a public meeting in the vicinity of the facility if local interest in the decommissioning of the facility is high. When the NRC holds a meeting with a licensee, members of the public are allowed to observe the meeting (except when the discussion involves proprietary, sensitive, safeguards, or classified information). The public has an opportunity to provide comments and there is an opportunity to request a hearing whenever a licensee submits a license amendment request, including an LTP or partial site release request.

The NRC would maintain the requirement to hold a public meeting upon submission of an LTP and a partial site release request. This proposed rule would modify the requirement for the NRC to hold a public meeting for the submittal of a PSDAR, LTP, or a partial site release request “in the vicinity of the site,” as currently required by §§ 50.82(a)(4)(ii), 50.82(a)(9)(iii), 50.83(f), 52.110(d)(2),52.110(i)(3), 52.111(f), 53.1070(d)(2), 53.1070(i)(3), and 53.1080(f). Since the COVID-19 pandemic, the NRC has seen an increase in the availability and accessibility of online platforms to conduct engaging, productive meetings with a variety of stakeholders from across the United States. In many cases, these meetings are able to reach more interested parties because they are more accessible than gathering at one physical location. However, the NRC recognizes that certain impacted communities, especially based on the use cases being presented to the Commission, may lack communication infrastructure that permits meaningful virtual participation. In such cases an in-person component to the public meeting would be necessary. The proposed modification of the locality requirement for the NRC to hold the public meetings associated with PSDARs, LTPs, and partial site release requests would modernize NRC regulations to conform to the ways the NRC currently conducts other public meetings while retaining an explicit option to have the public meeting in the vicinity of the facility, especially for cases with high local interest and/or less robust communication infrastructure. The proposed revision would provide the NRC with the explicit flexibility to hold these meetings in-person, virtually, or in a hybrid environment, enabling public participation and maintaining transparency while adjusting to the local interest and communication infrastructure. In addition, the format of the public meeting related to the licensee's request has no bearing on the risk to public health and safety from the licensee's request itself. Therefore, modifying the locality requirement would not present undue risk to public health and safety nor the common defense and security. The underlying purpose of the requirement is to provide for public participation in the NRC's consideration of the PSDAR, LTP, or partial site release, and this proposed rule would maintain the requirement for the NRC to hold a public meeting upon receipt of the PSDAR, LTP, or partial site release request. The proposed modification of the locality requirement, specifically, would apply to all required public meetings in the decommissioning process for power reactors, namely during NRC review of the PSDAR, LTP, and a partial site release request.

D. Shortening the Waiting Period Before Major Decommissioning Activities May Begin

Currently, a power reactor licensee may not begin major decommissioning activities until 90 days after the NRC receives the PSDAR and until certifications of permanent cessation of operations and permanent removal of fuel from the reactor vessel have been submitted, per §§ 50.82(a)(5), 52.110(e), and 53.1070(e). The purpose of this requirement is to provide the NRC with sufficient time to review the PSDAR before the licensee begins dismantlement activities that may increase risk onsite. When the regulation was established (61 FR 39301), the NRC noted that the 90-day waiting period does not address a health and safety issue but rather is the minimum time necessary for the NRC to evaluate the licensee's proposed activities and to conduct a public meeting. Notably, power reactor licensees have the authority to perform certain decommissioning activities during operation and throughout the decommissioning process in accordance with § 50.59 or § 53.1540. However, licensees are restricted from major decommissioning activities during this 90-day waiting period.

This proposed rule would reduce NRC resources dedicated to non-safety significant items related to the PSDAR, including responding to public comments submitted through the Office of the Federal Register. By removing the public notice and comment requirements, as discussed in section C, “Public Engagements Related to Decommissioning Licensing Actions,” this proposed rule would allow the staff to focus on risk-significant aspects of the review and shorten the review time without affecting safety. As a result, NRC review of the PSDAR would be completed within 30 days of receipt. Therefore, this proposed rule would reduce the waiting period from 90 days to 30 days. This proposed rule would make conforming changes to this requirement and others related to the submittal of the PSDAR to accommodate licensees that choose to submit an upfront LTP.

E. Minor Editorial and Administrative Corrections

This proposed rule includes the correction of a typographical error in § 50.59(b). That regulation currently applies, in part, to each holder of a combined license issued under 10 CFR part 52. However, the regulation incorrectly references the requirement for facilities licensed under 10 CFR part 52 to submit the certification of permanent cessation of operations as § 50.110. This rule proposes to amend this reference to the correct regulation, § 52.110(a). Additionally, this proposed rule would edit various requirements to harmonize regulatory language across current reactor licensing frameworks in 10 CFR parts 50, 52, and 53 to ensure consistent application of relevant decommissioning requirements.

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

A. Conforming the Financial Qualifications Requirements for 10 CFR Parts 50 and 52

Applicants With the Financial Qualifications Requirements for 10 CFR Part 53 Applicants

Under §§ 50.33(f) and 52.77, applications for utilization facility permits and licenses under 10 CFR parts 50 and 52, respectively, except for an electric utility applicant for a license to operate a utilization facility of the type described in § 50.21(b) or § 50.22, are required to include information sufficient to demonstrate that the applicant “possesses or has reasonable assurance of obtaining” the funds necessary for construction and/or operation, as applicable, of the facility along with associated financing details. However, applicants for commercial nuclear plants under the recently-issued 10 CFR part 53 are subject to the less restrictive financial qualifications standard of “appears to be financially qualified.” Part 53 is an optional ( printed page 60730) framework for the licensing of nuclear plants that could also be licensed under 10 CFR part 50 or 10 CFR part 52; therefore, the financial qualifications standard for 10 CFR part 53 applicants is applicable to 10 CFR parts 50 and 52 applicants. Having this less restrictive standard for 10 CFR parts 50 and 52 applicants would also reduce an unnecessary burden on these applicants and the NRC. Finally, having the same financial qualifications standard for 10 CFR parts 50, 52, and 53 applicants would further regulatory consistency. Therefore, the NRC is proposing to replace the current “possesses or has reasonable assurance of obtaining the funds necessary” financial qualifications standard in § 50.33(f)(1) and (2) with the 10 CFR part 53 “appears to be financially qualified” financial qualifications standard. And, as with 10 CFR part 53, the NRC is proposing to add to § 50.33(f)(1) and (2) that if available funding at the time of application is 50 percent or less, the applicant should include proposed license conditions to facilitate verification that funding is available prior to the start of construction and/or operation, as applicable. Making this conforming change to use the 10 CFR part 53 financial qualifications standard for 10 CFR parts 50 and 52 applicants is sufficiently protective of the public health and safety for the same reasons that the use of the standard was found to be sufficiently protective of the public health and safety for 10 CFR part 53 applicants.

B. Decommissioning Trust Fund Use

The NRC is proposing to delete §§ 50.82(a)(8)(ii), 52.110(h)(2), and 53.1045(a)(2) to remove the “timing” regulation that currently prohibits the use of decommissioning trust funds (DTFs) until decommissioning has begun ( e.g. the licensee has submitted a site-specific cost estimate and certifications of permanent cessation of operations and removal of fuel). Currently, a licensee may use up to 3% of the formula amount of its decommissioning funds (under § 50.75(c)) at any time for decommissioning planning, which includes paperwork activities and prohibits physical dismantlement and decontamination activities, without notification pursuant to §§ 50.75(h)(1)(iv), 50.75(h)(2), 53.1045(b)(4), and 50.1045(c). Currently, once certifications of permanent cessation of operations and fuel removal have been received and commencing 90 days after receipt of the PSDAR, an additional 20% of the DTF may be used for radiological decommissioning. The remaining funding may be used for radiological decommissioning once the site-specific cost estimate has been received. The proposed deletion of §§ 50.82(a)(8)(ii), 52.110(h)(2), and 53.1045(a)(2) would result in sole reliance on §§ 50.75(h)(1)(iv), 50.75(h)(2), 53.1045(b)(4), and 53.1045(c) for determining when and how decommissioning trust funds can be used. Therefore, the NRC is proposing to revise §§ 50.75(h)(1)(iv), 50.75(h)(2), 53.1045(b)(4), and 53.1045(c) to encompass similar timing regulations that define when and how much decommissioning trust funds may be used without notification to the NRC.

Proposed revisions to these regulations would allow the DTF to be used for legitimate radiological decommissioning activities during operations, typically for disposal of major radioactive components. Reviews of planned DTF withdrawals for these activities would fall under the notification process described in §§ 50.75(h)(1)(iv), 50.75(h)(2), 53.1045(b)(4), and 53.1045(c), rather than requiring an exemption from § 50.82, § 52.110, or § 53.1045 to facilitate the withdrawal. The proposed revision would still allow 3% of the formula amount to be used without notification and would also decrease the time the funds may be used once in decommissioning from 90 days to 30 days after the site-specific decommissioning cost estimate has been received, consistent with the proposed changes to §§ 50.82(a)(4), 52.110(d)(1), and 53.1070(d)(1).

The NRC is further proposing to revise §§ 50.75(h)(1)(iv), 50.75(h)(2), 53.1045(b)(4), and 53.1045(c) to require a 60-working day notification for such withdrawals. The current regulation requires a 30-working day notification during which the NRC receives and reviews the merits of the withdrawal. If the NRC does not object to the withdrawal, no further action is taken by the NRC (negative consent), and the DTF disbursement may occur after the 30-working days. Because withdrawals during operations for radiological decommissioning activities are not typical and administrative withdrawals are permitted without notification, the NRC does not typically receive these requests under the current regulatory structure. Although the current exemption process takes between 6-12 months to complete due to the development of supporting documents and evaluation of the request against the exemption criteria, the NRC typically completes the technical review to support withdrawals from decommissioning trust funds for major radioactive components within 60 days. Additionally, based on its experience evaluating previous exemptions, the NRC staff has found no issues of safety or security significance that would preclude a withdrawal of decommissioning trust funds during operations, that could not otherwise be determined during a standard review associated with the notification process.

Therefore, this proposed rule would include a 60-working day notification to allow for a complete technical analysis of a DTF withdrawal during operations. This proposed change in approach for how the NRC evaluates DTF use during operations would result in a net-positive time efficiency compared to the existing exemption process. The NRC has experience reviewing exemption requests for this purpose and, consistent with the proposed regulatory enhancements to §§ 50.75(h), 50.82(a)(8), 52.110(d)(1), 53.1045(b)(4), 53.1045(c), and 53.1070(d)(1), has determined that a detailed, streamlined review can be completed within 60 days, providing sufficient time to review the merits of the withdrawal and object, if necessary. This proposed change would increase overall efficiency for licensees to access decommissioning funds during operations by reducing regulatory and administrative burden and focusing NRC review on the technical merits of DTF withdrawals during operations, without a reduction in safety or security.

10 CFR Part 50—Risk Informed Seismic Design

The proposed changes to appendix S to 10 CFR part 50 would allow for the use of multiple design-basis ground motions (DBGMs). These changes would include:

a. A definition of DBGMs added under section III of appendix S to 10 CFR part 50;

b. A revised definition under section III of appendix S to 10 CFR part 50 related to the SSCs required to withstand the effects of vibratory ground motion or surface deformation. The revised definition would address both the SSE and DBGMs and the determination of SSCs based on their risk-significance and safety functions. Consistent changes would be made where the SSCs are referred to under part IV of appendix S to 10 CFR part 50;

c. A revised part IV, section (a)(1)(i) of appendix S to 10 CFR part 50 that would address the development of both the SSE and DBGMs and that would eliminate the minimum required peak ground acceleration for the response spectrum at the foundation level of structures. In place of the latter, and ( printed page 60731) consistent with 10 CFR part 53 requirements, the revised part IV, section (a)(1)(i) of appendix S to 10 CFR part 50 would enable the development of the response spectrum at the foundation level of structures commensurate with the risk-significance and safety functions of SSCs. Elsewhere in part IV of appendix S to 10 CFR part 50, references to SSE would be revised to include both the SSE and DBGMs;

d. A revised part IV, section (a)(1)(iv) of appendix S to 10 CFR part 50 to clarify that soil-structure interaction effects should be considered in the design of SSCs, if applicable. Further, part IV, section (a)(1)(iv) of appendix S to 10 CFR part 50 would be revised to replace the phrase “strain limits in excess of yield strain” with “inelastic behavior”;

e. A revised part IV, section (a)(2)(i)(A) and (B) of appendix S to 10 CFR part 50 that would define the operating basis earthquake based on the SSE or minimum DBGM; and

f. Other general proposed changes including removal of the start date for the applicability of appendix S to 10 CFR part 50, removal of a reference to appendix A to 10 CFR part 100, correcting a reference to § 50.10(c), replacing references to § 100.23 with references to 10 CFR part 100, and use of acronyms throughout appendix S to 10 CFR part 50.

10 CFR Part 50—Probabilistic Risk Assessment Requirements

The NRC proposes to amend § 50.69 to allow holders of an LWR CP under 10 CFR part 50, as well as applicants for an LWR design certification and holders of an LWR COL or ML under 10 CFR part 52, to make use of this regulation. Section 50.69 allows certain entities to categorize SSCs according to safety significance and, for safety-related SSCs of low safety significance, the ability to adopt alternatives to certain special treatment requirements that are in addition to the customary treatment of SSCs in industrial applications. The regulations currently permit this for holders of an LWR OL under 10 CFR part 50, holders of a renewed LWR license under 10 CFR part 54, applicants for a CP or an OL under 10 CFR part 50, and applicants for an SDA, a COL, or an ML under 10 CFR part 52. The NRC proposes to amend this regulation to extend its applicability to others. Specifically, § 50.69 would be available to holders of an LWR CP under 10 CFR part 50, as well as to applicants for an LWR design certification and holders of an LWR COL or ML under 10 CFR part 52. In addition, the NRC is proposing to clarify § 50.69(b) to make the applicability of the regulation more explicit.

This proposed change would allow for a risk-informed development and review of a design certification application. This would provide a safety benefit by permitting the applicant and the NRC to focus their efforts on the most risk-significant SSCs during design and design review. Risk-informed classification of SSCs in the design phase also would allow the use of alternative special treatment requirements for establishing the suitability of SSCs for the proposed design. Further, allowing LWR OL 10 CFR part 50 holders but not LWR COL 10 CFR part 52 holders to apply § 50.69 lacks a technical basis. Once the operating phase begins, there is no longer any technical difference between an OL and a COL. In addition, the proposed rule expands § 50.69 to ML holders to further provide flexibility to utilize risk-informed categorization and special treatments. This change would also be consistent with the Commission's PRA Policy Statement (60 FR 42622), which calls for increased use of PRA in all regulatory matters to the extent supported by the state of the art in PRA.

In SRM-SECY-18-0106, “Consideration in the Rulemaking Process of Issue Raised in Petition for Rulemaking on Applicability of Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Reactors (PRM-50-110) (NRC-2015-0028),” dated September 10, 2020 (84 FR 39684), the Commission directed the NRC staff to consider in the rulemaking process the issue of whether to allow holders of a COL to adopt risk-informed classifications of SSCs under § 50.69. These proposed changes would address that Commission direction.

10 CFR Parts 50, 52, 53, 70, and 72—Backfitting and Issue Finality Provisions

A. Compliance Backfitting

As part of this rulemaking, the NRC is proposing to remove the compliance backfit justifications. The compliance exception is intended to address situations in which the license failed to meet known and established standards of the Commission because of omission or mistake of fact. This exception was further clarified in SRM-COMSECY-16-0020. Consistent with the NRC's Principles of Good Regulation, removing the compliance backfit justifications would improve the efficiency, clarity, and reliability of the NRC's regulations. With respect to efficiency, the Principles of Good Regulation state that regulatory activities should be consistent with the degree of risk reduction they achieve. Adequate protection backfits and backfits involving a cost-justified substantial increase in overall protection inherently apply clear and reliable standards that utilize the NRC's resources in a manner commensurate with an activity's associated risks. In contrast, the compliance backfit provisions do not provide a specific threshold for evaluating the degree of risk-reduction, and in so doing, are susceptible to focusing resources on less risk-significant issues. Thus, in applying the compliance exception, a licensee may be required to make a change without a showing of a substantial safety benefit. Further, while the NRC's implementing guidance directs “some consideration of cost” in its documented evaluation for a compliance backfit ( e.g., MD 8.4), the backfitting and issue finality regulation provisions do not explicitly specify that cost must be considered for compliance backfits. Moreover, due to the lack of specific criteria in the compliance backfit provisions, the compliance justifications are susceptible to less objective application and, therefore, are less predictable and may create some regulatory uncertainty for entities when making long-term financial decisions regarding, for example, new facility construction or major upgrades. Such uncertainty may create a barrier to competition; therefore, in response to E.O.14627, the NRC is proposing to eliminate the compliance backfit provisions from its rules.

In addition, the backfitting and issue finality rules in 10 CFR parts 50, 52, and 53, and the Commission's policy described in MD 8.4, require the NRC to demonstrate that certain criteria are met before imposing new or revised requirements that would result in a change to the design, construction, or operation of a nuclear power reactor. These regulations apply upon the issuance of certain permits, licenses, certifications, or other approvals under 10 CFR parts 50, 52, and 53, including the issuance of a construction permit for a power reactor under 10 CFR part 50 or 10 CFR part 53 or a combined construction and operating license issued under 10 CFR part 52 or 10 CFR part 53. Accordingly, the proposed removal of the compliance backfit justification from §§ 50.109 and 53.1590 and from the issue finality provisions in 10 CFR parts 52 and 53 would further the direction in section 5(f) of E.O. 14300 to provide stringent thresholds for circumstances in which the NRC ( printed page 60732) may demand changes to a reactor design once construction is underway.

Finally, the issue finality regulations in §§ 52.63(a)(4) and 53.1263(a)(4) state that the Commission may not impose new requirements by plant-specific order on any part of the design of a specific plant referencing a design certification rule if that part was approved in the design certification, unless (1) a modification is necessary for adequate protection or compliance and (2) special circumstances as defined in § 52.7 or § 53.080 are present and outweigh any decrease in safety that may result from a reduction in standardization. There are similar provisions in paragraph VIII.B.3 of the design certification rules located in the appendices to 10 CFR part 52. With the proposed removal of the compliance justification, it would not be necessary for the NRC to separately demonstrate that special circumstances are present when applying § 52.63(a)(4), § 53.1263(a)(4), or paragraph VIII.B.3 of the design certification rules. That is, when such a modification is necessary for the NRC to have reasonable assurance of adequate protection, the AEA requires the NRC to take regulatory action regardless of the presence of special circumstances. Therefore, this proposed rule would remove the special circumstances criteria from §§ 52.63(a)(4), 53.1263(a)(4), and paragraph VIII.B.3 of the design certification rules in appendices A, D, E, F, and G to 10 CFR part 52. This proposed rule would also correct an error in § 53.1263(a)(4) by replacing the reference to “§ 53.1248” with “§ 53.1251 or § 53.1260.”

B. Applicability of the Backfitting and Issue Finality Regulations in 10 CFR Parts 50 and 52

This proposed rule would amend §§ 50.109, 52.63, and 52.171 to revise and clarify the applicability of the backfitting and issue finality requirements in 10 CFR parts 50 and 52. The backfitting and issue finality requirements in 10 CFR parts 50 and 52 overlap in some areas and have inconsistencies that may lead to confusion about the applicable criteria for the NRC to impose certain changes. The applicability of § 50.109 would also be updated to reflect the Commission's interpretation and policies related to its application and to provide more consistency regarding the backfitting of nuclear power plants licensed under 10 CFR parts 50, 52, and 53. This proposed rule would also remove certain provisions regarding the applicability of § 50.109 that are obsolete or unnecessary.

Currently, the licenses, permits, and approvals that are within the scope of § 50.109 are listed in § 50.109(a)(1)(i) through (vii) under the definition of backfitting. The NRC is proposing to clarify this list and relocate it to a new paragraph (a) that describes the applicability of § 50.109. The existing paragraph (a) in § 50.109 would replace the deleted paragraph (b), which is no longer necessary. The proposed § 50.109(a)(1) would identify the specific approvals issued under 10 CFR parts 50 and 52 for which § 50.109 is applicable. Additionally, the proposed § 50.109(a)(2) would clarify that matters addressed in a design certification or manufacturing license referenced in a construction permit, operating license, or combined license for a power reactor issued under 10 CFR part 50 or 10 CFR part 52 are not within the scope of § 50.109 but instead must be considered under the appropriate issue finality requirements in 10 CFR part 52.

The regulations in 10 CFR part 50 provide for the licensing of production and utilization facilities in general, and not just nuclear power reactors. The NRC has always considered nuclear power reactors licensed under section103 or 104b of the AEA ( i.e., commercial nuclear power reactors) to be within the scope of § 50.109. In 2026 (91 FR 2287), the NRC issued a notice of interpretation explaining that commercial NPUFs licensed under section 103 of the AEA are included within the scope of § 50.109 and that non-commercial NPUFs licensed under section 104 of the AEA are excluded from the scope of § 50.109. Licenses issued under section 103 of the AEA are referred to as class 103 licenses in the Commission's regulations in 10 CFR part 50, and are issued under § 50.22. In addition, licenses, permits, and other Commission authorizations issued under 10 CFR part 53 are not within the scope of § 50.109. Therefore, the proposed amendment of § 50.109 would clarify in the new paragraph (a)(1) that, among other things, § 50.109 applies after the date of issuance of the following:

a. Construction permits and operating licenses for nuclear power reactors issued under 10 CFR part 50;

b. SDAs for nuclear power reactors issued under subpart E of 10 CFR part 52;

c. Combined licenses for nuclear power reactors issued under subpart C of 10 CFR part 52; and

d. Construction permits and operating licenses for commercial NPUFs issued under § 50.22.

The proposed § 50.109(a)(2) would further clarify that the NRC would apply § 50.109 to matters within the scope of these approvals, which include renewed, amended, or revised approvals, except for matters resolved in a design certification rule or manufacturing license proceeding referenced in a construction permit, operating license, or combined license for a power reactor issued under 10 CFR part 50 or 10 CFR part 52.

The NRC is also proposing to add limited work authorizations issued under § 50.10 to the scope of § 50.109 for nuclear power reactors licensed or to be licensed under 10 CFR part 50 or 10 CFR part 52 and for commercial NPUFs licensed or to be licensed under § 50.22. The Commission's policy described in MD 8.4 states that the NRC would apply § 50.109 to matters covered by a limited work authorization for power reactors to be licensed under 10 CFR part 50. Since MD 8.4 was issued in 2019, the NRC issued an interpretation that commercial NPUFs licensed under § 50.22 are within the scope of § 50.109 (91 FR 2287). The regulations in § 50.10 apply to the issuance of all limited work authorizations under 10 CFR parts 50 and 52, and § 50.10 does not provide different standards for power reactors to be licensed under 10 CFR part 50 or 10 CFR part 52 or commercial NPUFs to be licensed under § 50.22. In addition, the backfitting rule in § 53.1590 applies to limited work authorizations for nuclear power reactors licensed or to be licensed under 10 CFR part 53. To provide consistent treatment under the backfitting provisions across frameworks, the proposed § 50.109(a)(1) would add limited work authorizations for nuclear power reactors and commercial NPUFs licensed or to be licensed under 10 CFR part 50 or 10 CFR part 52, as applicable, to the scope of § 50.109.

In addition, the current provisions in § 50.109(a)(1)(i) and (ii) provide a distinction between the applicability of the backfitting provisions for nuclear power reactor construction permits issued before and after October 21, 1985. There are no longer any nuclear power reactor construction permits in effect that were issued before October 21, 1985. The proposed amendment would remove § 50.109(a)(1)(i) for construction permits issued before October 21, 1985, and remove the date from § 50.109(a)(1)(ii) (and renumber the provision) because it is no longer necessary to provide a distinction regarding when a construction permit was issued. Similarly, the NRC also proposes to remove § 50.109(b) because it is no longer necessary.

The NRC is proposing to delete § 50.109(a)(1)(vi) because it is redundant ( printed page 60733) to § 50.109(a)(1)(i). The current provision in § 50.109(a)(1)(vi) makes § 50.109 applicable after the date of issuance of the first construction permit issued for a duplicate design ( i.e., identical design) for a nuclear power reactor under appendix N, “Standardization of Nuclear Power Plant Designs: Permits to Construct and Licenses to Operate Nuclear Power Reactors of Identical Design at Multiple Sites,” to 10 CFR part 50. Appendix N to 10 CFR part 50 concerns applications “filed by one or more applicants for licenses to construct and operate nuclear power reactors of essentially the same design to be located at different sites.” The issuance of a construction permit is the only method provided in appendix N to CFR part 50 for the approval of a duplicate design that may be referenced by other construction permit applications. This process would allow for one review—at the first construction permit application stage—of those aspects of the design that are approved and then referenced in subsequent appendix N to 10 CFR part 50 applications. As discussed in the preamble for the 2007 10 CFR part 52 final rule, § 50.109(a)(1)(vi) was added to clarify when the applicability of § 50.109 begins for a construction permit for a duplicate design. However, the current provision in § 50.109(a)(1)(i), which will be retained in the proposed § 50.109(a), makes § 50.109 applicable to all construction permits issued for nuclear power reactors under 10 CFR part 50, which would include the first construction permit issued for a duplicate design issued under appendix N to 10 CFR part 50. Accordingly, when the first construction permit is issued under appendix N, then § 50.109 becomes applicable on that date for all subsequent appendix N construction permit applications that reference the design of the first construction permit (and the construction permits themselves if they are issued). Therefore, the current provision in § 50.109(a)(1)(vi) is not necessary.

The current provision in § 50.109(c)(2) would be revised to require a backfit analysis to also include a general description of the activities that would be required by holders of SDAs to implement a backfit. Currently, § 50.109(c)(2) only requires a general description of the activities required by licensees and applicants. Under the definitions of “License,” “Licensee,” and “Standard design approval” in § 52.1(a), a holder of an SDA is not a licensee.

The NRC also proposes to remove manufacturing licenses from the scope of the backfitting rule in § 50.109 to eliminate inconsistences between §§ 50.109, 52.98, and 52.171. Currently, manufacturing licenses are included within the definition of backfitting in § 50.109(a)(1), and § 50.109(a)(1)(v) includes manufacturing licenses within the scope of § 50.109. However, this is inconsistent with other provisions in §§ 50.109, 52.98, and 52.171 that require the NRC to apply the issue finality criteria in § 52.171 to impose changes related to a manufacturing license including the design of a manufactured reactor. For example, § 52.171(a) requires the NRC to apply the criteria in § 52.171(a) “notwithstanding any provision in 10 CFR 50.109.” For a combined license that references a reactor manufactured under a manufacturing license, §§ 50.109(a)(1)(vii) and 52.98(d) currently specify that the provisions of § 52.171 apply with respect to matters resolved in the manufacturing license proceeding. This proposed rule would remove the words “or manufacturing license” from the definition of backfitting in § 50.109(a)(1) and remove § 50.109(a)(1)(v) to resolve this inconsistency. As a result, NRC-imposed changes to manufacturing licenses would be governed by the provisions in § 52.171. This proposed rule would also delete the phrase “notwithstanding any provision in 10 CFR 50.109” from § 52.171(a) because the proposed changes to clarify § 50.109 would make this clause unnecessary.

For a combined license, § 50.109(a)(1)(vii) currently clarifies that the issue finality regulations in § 52.63, § 52.171, or the referenced design certification rule apply to matters resolved in referenced standard design certification rule or manufacturing license proceeding. In addition, § 50.109(a)(1)(vii) currently clarifies that any backfitting limitation in a referenced standard design certification rule would govern over the provisions in §§ 50.109 and 52.63. However, the provisions in § 50.109(a)(1)(i) and (iii) that incorporate construction permits and operating licenses, respectively, within the scope of § 50.109 do not include these clarifications. To ensure consistency in the application of § 50.109, the clarification in § 50.109(a)(1)(vii) regarding the applicability of these issue finality requirements would be relocated to the proposed § 50.109(a)(2) to provide clarification for power reactor construction permits, operating licenses, and combined licenses issued under 10 CFR parts 50 and 52. Specifically, the proposed § 50.109(a)(2) would clarify that the provisions of § 52.63, § 52.171, or the referenced design certification rule apply to matters resolved in a referenced standard design certification rule or manufacturing license proceeding, as appropriate, and § 50.109 would not apply. The proposed § 50.109(a)(2) would also state that any backfitting limitation in a referenced standard design certification rule would govern over the provisions in §§ 50.109 and 52.63.

The NRC proposes to delete the phrase “notwithstanding any provision in 10CFR 50.109” from § 52.63(a)(1) to avoid potential inconsistencies with other rules. The regulation in § 52.63(a)(1) states that, “notwithstanding any provision in 10 CFR 50.109,” changes to certification information must be made in accordance with the criteria in § 52.63(a)(1). The design certification rules codified in appendices to 10 CFR part 52 that are currently in effect include provisions that specify the information subject to § 52.63(a) and the information subject to § 50.109. For example, paragraph VIII.C.1 in appendix D to 10 CFR part 52 states, in part, “Generic changes to generic TS [technical specifications] and other operational requirements that were completely reviewed and approved in the design certification rulemaking and do not require a change to a design feature in the generic DCD [design control document] are governed by the requirements in 10 CFR 50.109.” In addition, the proposed § 50.109(a)(2) would clarify the applicability of § 52.63 and the design certification rules if they are referenced in a power reactor license or construction permit issued under 10 CFR part 50 or 10 CFR part 52, as applicable. Therefore, the “notwithstanding” clause in § 52.63(a)(1) is not necessary. In addition, eliminating this clause would clarify that the provisions in the design certification rules establish the appropriate change control mechanisms for different types of information associated with the design certification.

Next, to enhance clarity and consistency across NRC regulations, the NRC is proposing to delete the sentence in § 50.109(a)(1)(vii) that states that, for a combined license referencing an SDA, § 52.145 applies to the design matters resolved in that SDA. The NRC proposes to remove this text because §§ 52.98 and 52.145 and the associated discussions in the preamble for the 2007 10 CFR part 52 final rule (72 FR 49352) explain that issue finality under § 52.145 applies only when the NRC staff and Advisory Committee on Reactor Safeguards are reviewing an individual facility license application incorporating an SDA. In ( printed page 60734) contrast, the Commission, Atomic Safety Licensing Board Panel, or presiding officers are not bound by NRC staff determinations in the final design approval or final safety evaluation report for the SDA. The NRC also proposes to delete the sentence in § 50.109(a)(1)(vii) stating that § 52.145 applies to the design matters resolved in the SDA because the existing text could unintentionally imply a conflict with the issue finality requirements in § 52.98 for combined licenses referencing SDAs. Specifically, § 52.98(a) requires NRC-initiated changes to a combined license related to design matters resolved in an SDA to be addressed under § 50.109, not § 52.145. This point is reinforced by the Commission's discussion for § 52.98 in the preamble for the 2007 10 CFR part 52 final rule, which states, in part, that “[t]he change processes in 10 CFR part 50 apply to a combined license that does not reference a design certification rule or a reactor manufactured under a manufacturing license.” As such, this would mean that the change processes in 10 CFR part 50 apply to combined licenses that reference an SDA. Thus, § 52.145 does not apply to a combined license that references an SDA after the license is issued.

As previously stated, the issue finality provisions of § 52.145 address the finality of an SDA in the review of an application referencing the SDA. Under current regulations, NRC-imposed generic changes to an SDA are made under § 50.109, as reflected by the fact that standard design approvals are included within the definition of backfitting in § 50.109(a)(1), and § 50.109(a)(1)(iv) establishes the date when an SDA is included within the scope of § 50.109. These requirements in § 50.109 would be retained to address generic NRC-imposed changes to an SDA, but the requirements in § 50.109(a)(1)(iv) would be relocated to the proposed § 50.109(a)(1).

The NRC also proposes to delete the sentence in § 50.109(a)(1)(vii) stating that, for a combined license referencing an early site permit, § 52.39 applies to the site characteristics, design parameters, and terms and conditions specified in the early site permit after the combined license is issued. The NRC proposes to remove this text because it is incorrect. Section 52.26(d) states that upon issuance of a construction permit or combined license, a referenced early site permit is subsumed, to the extent referenced, into the construction permit or combined license. Therefore, an early site permit is no longer operative once the construction permit or combined license is issued, so § 52.39 cannot apply to an issued construction permit or combined license that references an early site permit.

The NRC proposes to amend § 52.171(a)(1) to clarify that restriction on NRC-imposed changes related to the design of a nuclear power reactor approved under a manufacturing license applies before, during, and after the reactor is manufactured. The current wording of § 52.171(a)(1) could be read to suggest that this restriction applies only when the reactor is being manufactured. Additionally, § 52.171(a)(1) indicates that this regulation only applies during the term of the manufacturing license. However, such a reading would not be consistent with the preamble for the 2007 10 CFR part 52 final rule. The Commission explained in the preamble that under § 52.171(a)(1) “the NRC may not impose any change or modification to the approved design (including site parameters, or design characteristics) for the manufacturing license” unless the NRC determines that the § 52.171(a)(1) criteria for imposing the change are met. The Commission further explained that this restriction applies “to reactors which have already been manufactured, as well as any reactors yet to be manufactured under the manufacturing license.”

Accordingly, the NRC is proposing to revise § 52.171(a)(1) to more clearly reflect the intent of the regulation as promulgated in the 2007 10 CFR part 52 final rule. With the proposed changes, the design of a manufactured reactor referenced in an operating license, for example, would remain within the scope of § 52.171(a)(1) independent of the duration of the manufacturing license. With respect to the manufacturing license itself, it is not necessary to specify in § 52.171(a)(1) that this provision applies “during the term of the manufacturing license” because, even without this clause, an expired manufacturing license would not be within the scope of § 52.171(a)(1).

C. Applicability of the Backfitting and Issue Finality Regulations in 10 CFR Part 53

In general, the backfitting and issue finality provisions in 10 CFR part 53 are similar to the backfitting and issue finality provisions in 10 CFR parts 50 and 52. The proposed amendments to §§ 53.1288, “Finality of manufacturing licenses,” and 53.1590 would revise and clarify the applicability of the backfitting and issue finality provisions in 10 CFR part 53. Many of these proposed changes are, in part, intended to align with §§ 50.109 and 52.171, where practical. The backfitting and issue finality regulations in 10 CFR part 53 overlap in some areas, which may lead to confusion about the applicable criteria for the NRC to impose certain changes.

Currently, § 53.1590 applies after the date of issuance of a commercial nuclear plant license issued under 10 CFR part 53. This means that the scope of § 53.1590 includes construction permits, operating licenses, combined licenses, and limited work authorizations for nuclear power plants licensed under 10 CFR part 53 based on the definitions for “commercial nuclear plant” and “license” in § 53.020, “Definitions.” However, § 53.1590 does not specify how the issue finality rules are to be applied when a license references a design certification rule or manufacturing license. In addition, the definition of backfitting in § 53.1590 includes “design approvals,” which is equivalent to the term “SDA” according to the definition in § 53.020. But SDAs are not a license under 10 CFR part 53 and, as a result, it is not clear whether an SDA could fall within the scope of § 53.1590.

The NRC is proposing to add a new paragraph (a) to § 53.1590 to clarify the applicability of § 53.1590 and the related issue finality regulations. The existing paragraphs (a) through (d) in § 53.1590 would be redesignated as paragraphs (b) through (e). The proposed § 53.1590(a)(1) would identify that § 53.1590 applies after construction permits, operating licenses, combined licenses, limited work authorizations, and SDAs are approved under 10 CFR part 53. The proposed § 53.1590(a)(2) would clarify that matters addressed in a design certification or manufacturing license referenced in such NRC approvals under 10 CFR part 53 are not within the scope of § 53.1590 but instead must be considered under the appropriate issue finality requirements in 10 CFR part 53.

The proposed inclusion of SDAs within the scope of § 53.1590 would be consistent with § 50.109, which includes SDAs issued under 10 CFR part 52 within its scope. The issue finality regulations in § 53.1221 would continue to apply for license applications under 10 CFR part 53 that reference an SDA. Applications referencing an SDA would remain outside the scope of the backfitting provisions in § 53.1590. However, the NRC would apply the provisions in § 53.1590 when considering generic changes that would affect an SDA.

The proposed § 53.1590(a)(2) would specify that the provisions of § 53.1263, § 53.1288, or the referenced design ( printed page 60735) certification rule apply—and § 53.1590 would not apply—to matters resolved in a referenced standard design certification rule or manufacturing license proceeding. The proposed § 53.1590(a)(2) would also state that any backfitting limitation in a referenced standard design certification rule would govern over the provisions in §§ 53.1590 and 53.1263. The proposed § 53.1590(a)(2) provisions are equivalent to the proposed § 50.109(a)(2) and the provisions in § 50.109(a)(vii) for combined licenses issued under 10CFR part 52 that reference a design certification rule or manufacturing license. For combined licenses issued under 10 CFR part 53, the proposed § 53.1590(a)(2) is consistent with the issue finality provisions in §§ 53.1425, “Finality of referenced NRC approvals,” and 53.1443, “Finality of combined licenses.” For example, § 53.1443(c)(1) states that if a combined license references a certified design, then “changes to or departures from information within the scope of the referenced standard design certification rule are subject to the applicable change processes in that rule.” Similarly, § 53.1443(d)(1) states that if a combined license references use of a manufactured reactor, then “changes to or departures from information within the scope of the manufactured reactor's design are subject to the change processes in § 53.1288.” However, like 10 CFR parts 50 and 52, the regulations in 10 CFR part 53 also allow construction permits and operating licenses to reference a design certification rule or a manufacturing license, but 10 CFR parts 50 and 52 do not provide issue finality provisions equivalent to §§ 53.1425 and 53.1443 that apply after the construction permit or operating license is issued. To ensure consistency between frameworks regarding the application of the backfitting and issue finality regulations, the proposed § 53.1590(a)(2) would apply to construction permits, operating licenses, and combined licenses issued under 10 CFR part 53 after such an approval is issued.

The current provision in § 50.1590(b)(2) would be revised to require a backfit analysis to also include a general description of the activities that would be required by holders of SDAs to implement a backfit. Currently, § 50.1590(b)(2) only requires a general description of the activities required by licensees and applicants.

The NRC proposes to delete the clause in § 53.1288(a)(1) implying that this provision regarding the issue finality for manufacturing licenses only applies during the term of a manufacturing license issued under this part. This proposed provision would apply to the design of manufactured reactors before, during, and after they have been manufactured under a manufacturing license. For example, § 53.1443(d)(1) states that if a combined license references use of a manufactured reactor, then “changes to or departures from information within the scope of the manufactured reactor's design are subject to the change processes in § 53.1288.” The regulation in § 53.1288(a)(1) was based on § 52.171(a)(1), and, as discussed previously, the current wording of § 52.171(a)(1) is inconsistent with the Commission's intent when the final 2007 10 CFR part 52 rule was issued. With the proposed changes, the design of a manufactured reactor referenced in an operating license, for example, would remain within the scope of § 53.1288(a)(1) independent of the duration of the manufacturing license. With respect to the manufacturing license itself, it is not necessary to specify in § 53.1288(a)(1) that this provision applies during the term of the manufacturing license because, even without this clause, an expired manufacturing license would not be within the scope of § 53.1288(a)(1).

Next, in accordance with § 53.620(d), “Fuel loading,” a manufacturing license under 10 CFR part 53 may authorize the possession and loading of fresh fuel in a manufactured reactor pursuant to a license issued under 10 CFR part 70. Among other requirements, § 53.620(d)(2) states that “holders of these part 70 licenses must comply with the requirements of Subpart H to part 70, regardless of whether their proposed activities meet the applicability criteria found in 10 CFR 70.60.” Although the licensee must comply with the requirements of subpart H to 10 CFR part 70, the backfitting provision in § 70.76 applies to the NRC and not the licensee. Additionally, the regulation in § 53.620(d)(1)(ii) allows an applicant for a manufacturing license to file a separate, subsequent application for a 10 CFR part 70 license or to combine both applications. As a result, the provision in § 53.620(d) and subpart H to 10 CFR part 70, create some uncertainty regarding the applicability of the backfitting rule in § 70.76 and the issue finality rule in § 53.1288(a).

To rectify this potential uncertainty, the NRC proposes adding a new paragraph § 53.1288(a)(4) to clarify the applicability of the issue finality provisions in § 53.1288 and the backfitting provisions in § 70.76 for manufacturing licensees that have obtained a 10 CFR part 70 license in accordance with § 53.620 to possess and load fresh fuel into a manufactured reactor. Specifically, the proposed § 53.1288(a)(4) would clarify that the 10 CFR part 70 license would be within the scope of the backfitting rule in § 70.76. The backfitting requirements in § 70.76 would be appropriate for NRC-initiated changes that may affect the 10 CFR part 70 license because the activities authorized under the 10 CFR part 70 license, such as the possession, storage, and handling of fresh fuel, are similar to activities authorized under other 10 CFR part 70 licenses. Additionally, the 10 CFR part 70 license would be subject to the requirements in subpart H to 10 CFR part 70, and may be issued separately or subsequent to the issuance of the manufacturing license. However, the features to prevent criticality in a manufactured reactor that are specified in the manufacturing license, as required by § 53.620(d)(1)(i), would be within the scope of the issue finality provisions in § 53.1288(a)(1) under this proposed rule because these features would have been subject to the manufacturing license proceeding and are expected to be integral to the design and manufacture of the reactor.

D. Other Revisions to Criteria for Changes Affecting Issue Finality

The NRC proposes to revise the criteria in §§ 52.63(a)(1)(iii) and 53.1263(a)(1)(iii) that allows the Commission to change requirements on design certification information if the change reduces unnecessary regulatory burden and maintains protection to public health and safety and the common defense and security. Specifically, the NRC proposes to change the phrase “maintains protection to” to “maintains reasonable assurance of adequate protection of.” The preamble discussion for § 52.63(a)(1)(iii) in the final 2007 10 CFR part 52 rule stated, in part, that “maintaining protection generally embodies the same safety principles used by the NRC in applying risk-informed decision-making, i.e., ensuring that adequate protection is provided . . . .”

As discussed in MD 8.4, the Commission's policy requires decisions regarding changes affecting issue finality to be risk-informed, which is consistent with the concept of “maintaining protection” discussed in the preamble for the final 2007 10 CFR part 52 rule. The proposed changes to §§ 52.63(a)(1)(iii) and 53.1263(a)(1)(iii) would clarify that the NRC must demonstrate that reasonable assurance of adequate protection would be maintained when making changes that reduce regulatory burden and would continue to allow the NRC to make risk- ( printed page 60736) informed decisions regarding such changes.

The NRC proposes adding criteria to the issue finality provisions in §§ 52.171(a) and 53.1288(a) that would allow the Commission to change requirements to reduce unnecessary regulatory burden on manufacturing licenses and licensees with manufactured reactors provided that the NRC continues to have reasonable assurance of adequate protection of the public health and safety and the common defense and security. The provisions in §§ 52.171(a) and 53.1288(a) currently provide that the Commission may not modify, rescind, or impose new requirements on the design or manufacture of a manufactured reactor unless the change can be justified as necessary to provide reasonable assurance of adequate protection or for compliance with NRC requirements. These provisions do not appear to allow the NRC to eliminate or relax requirements that are not necessary to provide reasonable assurance of adequate protection of public health and safety or the common defense and security. The proposed criteria to be added to §§ 52.171(a) and 53.1288(a) are equivalent to the criteria currently in §§ 52.63(a)(1)(iii) and 53.1263(a)(1)(iii) with the proposed clarifications previously discussed.

E. Clarification of Backfitting Definitions

The NRC proposes to revise the definitions of “backfitting” in §§ 50.109, 53.1590, and 70.76 to clarify that backfitting occurs when certain changes to the facility, design, or procedures “would be required by” a new or amended NRC requirement or staff position. Currently, the backfitting definitions state that backfitting occurs when certain changes to the facility, design, or procedures “may result from” a new or amended NRC requirement or staff position. The Commission's policy described in MD 8.4 identifies that nonmandatory relaxations of regulatory requirements or staff positions are not backfits. The phrase “may result from” could be interpreted as requiring the NRC to consider backfitting even for nonmandatory relaxations. Therefore, the NRC is proposing to clarify that definitions of “backfitting” in §§ 50.109, 53.1590, and 70.76 are limited to situations where the NRC action would require changes to the facility, design, or procedures consistent with the Commission's policy on nonmandatory relaxations.

10 CFR Parts 52 and 53—Notification of Initial Fuel Load

The NRC is proposing to revise certain requirements related to the notification of initial fuel load in §§ 52.103(a) and 53.1452(a). The proposed change to § 52.103(a) would eliminate the requirement for recurrent reporting of updates to the licensee's scheduled date for initial fuel load every 30 days after the initial report submitted no less than 270 days before the scheduled date. The licensee would be required to notify the NRC only in those cases where, after providing the initial notification, the scheduled date for initial fuel load is revised. In such cases, the licensee would be required to notify the NRC within 30 days of revising the schedule.

The proposed change would also be applicable to COL holders installing a fueled manufactured reactor under 10 CFR part 53. The proposed change would eliminate the requirement for recurrent reporting of updates to the COL holder's scheduled date for initiating the removal of the features to prevent criticality every 30 days after the initial report submitted no less than 270 days before the scheduled date. The COL holder would be required to notify the NRC only in cases where, after providing the initial notification, the scheduled date for initiating the removal of the features to prevent criticality is revised. In such cases, the COL holder would be required to notify the NRC within 30 days of revising the schedule.

These amendments would reduce unnecessary regulatory burden without any reduction in safety or security.

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

The following regulations concerning Unresolved Safety Issues, Generic Safety Issues, and operating experience in relation to plant designs are proposed to be removed for future new reactor applicants:

• § 52.47(a)(21) and (22) for design certifications under10 CFR part 52, subpart B
• § 52.79(a)(20) and (37) for combined licenses under10 CFR part 52, subpart C
• § 52.137(a)(21) and (22) for standard design approvals under10 CFR part 52, subpart E
• § 52.157(f)(28) and (29) for manufacturing licenses under10 CFR part 52, subpart F

The relevant regulatory guidance is currently located in Regulatory Guide 1.206 and Chapter 1 of NUREG-0800, both of which are for LWRs. There is a brief discussion concerning Unresolved Safety Issues, Generic Safety Issues, and operating experience in relation to plant designs for non-LWRs in NRC Interim Staff Guidance in DANU-ISG-2022-01, “Review of Risk-Informed, Technology-Inclusive Advanced Reactor Applications—Roadmap,” dated March 2024, (ML23277A139). There is no specific guidance for non-LWRs; the NRC would consider the existing LWR guidance to inform the NRC's licensing review. In association with the proposed removal of regulations, the NRC would revise the relevant guidance. The revised guidance would include a discussion related to applications under 10 CFR part 50.

The NRC recently issued Regulatory Guide 1.254, “Technology-Inclusive Identification of Licensing Events for Commercial Nuclear Plants,” March 2026, (ML25232A005) for nuclear plants under 10 CFR parts 50, 52, and 53. This document integrates operating experience as a foundational input in several key areas of the safety analysis process. Under the guidance, applicants are encouraged to systematically review operating experience from a variety of sources, including NRC communications, industry databases, and international reports.

The expected impact on future applicants and other stakeholders from the rulemaking would be a reduction in burden without compromising safety. Specifically, applicants would no longer be required to submit for NRC review an evaluation of USIs, GSIs, and operating experience from generic communications relevant to their design. The NRC would continue to consider operating experience of potential safety or risk significance during its review of future new reactors under other regulatory requirements and guidance that are sufficient to reach a finding of reasonable assurance of adequate protection of public health and safety. Another impact of this proposed rule would be that it would increase licensing efficiency and effectiveness because of the reduced content of applications and scope of licensing reviews, thus allowing resources to be allocated to other safety- or risk-significant issues.

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

A. Elimination of SDA Expiration

This proposed rule would eliminate the current 15-year expiration period and prohibition on renewal for future SDAs, as specified in §§ 52.147 and 53.1218. Under the revised framework, ( printed page 60737) once an SDA is approved by the NRC, it would remain valid indefinitely unless it is withdrawnby the applicant, supersededby a subsequent design approval or certification, or materially changedin a way that affects safety or regulatory compliance.

This change would align with the NRC's goal of reducing unnecessary regulatory burden and increasing regulatory stability for applicants and licensees. It recognizes that design information, particularly for advanced reactors, may remain technically relevant and safe for extended periods. Removing the expiration date would encourage long-term investment in design development and support deployment flexibility.

For an SDA issued following this proposed change, the holders would no longer need to reapply for approval solely due to the passage of time. However, the holder would still be responsible for ensuring the design remains technically current and safe, especially if used in future license applications. For an SDA issued prior to the proposed change, the SDA holder could choose to submit an amendment requesting the expiration date to be removed from the SDA, under the proposed § 52.145.

B. Inclusion of SDA in § 50.59, § 53.1221, and § 53.1505

This proposed rule would explicitly include SDAs within the scope of the §§ 50.59 and 53.1505, “Changes to licensing-basis information requiring prior NRC approval.” The change proposed in § 50.59 would allow 10 CFR part 52 SDA holders to make certain changes to the approved design without prior NRC approval, provided the changes do not result in more than a minimal increase in the frequency or consequences of accidents, create the possibility of a new or different kind of accident, reduce the margin of safety, or otherwise meet the criteria requiring prior NRC review.

For 10 CFR part 53, this proposed rule would also add a new paragraph (e) to § 53.1221, which would allow changes to SDAs that meet the criteria in § 53.1550(a)(2) and do not require a change to the associated generic technical specifications. This would allow SDA holders under 10 CFR part 53 to make certain changes to their approved designs without prior NRC approval, provided the changes do not:

a. Result in an increase to the frequency or consequences of an event sequence such that an event sequence not previously identified as risk significant becomes risk significant;

b. Result in an increase to the frequency or consequences of an event sequence such that an event sequence exceeds the licensing-basis event evaluation criteria required to be established in accordance with § 53.450(e);

c. Involve either of the following: (A) a change to the NRC-approved comprehensive risk metric(s) or associated risk performance objective under § 53.220(b), or (B) an increase to the frequency or consequences of one or more event sequences such that any calculated comprehensive risk metric exceeds the associated risk performance objective established in accordance with § 53.220;

d. Involve a departure from a method of evaluation described in the FSAR (as updated) used in assessing design-basis accidents in accordance with § 53.450(f) unless the results of the analysis under § 53.450(f) are conservative or essentially the same, the revised method of evaluation has been previously approved by the NRC for the intended application, or the revised method of evaluation can be used under an NRC-endorsed consensus code or standard;

e. Result in a change to the safety classification of an SSC from safety-related to either nonsafety-related but safety-significant or non-safety-related;

f. Result in more than a minimal decrease in defense in depth;

g. Result in the identification of a new design-basis accident in accordance with § 53.450(f) or

h. Result in more than a minimal increase in the consequences of any design-basis accident.

This change would provide consistency with how other licensing instruments ( e.g., facility operating licenses and certified designs) are treated under the NRC's regulatory framework and would empower SDA holders to manage minor design evolutions efficiently while maintaining the appropriate degree of NRC oversight.

SDA holders would also gain flexibility to implement design improvements or updates based on operational experience, technological advancements, or vendor feedback. SDA holders would still be required to maintain sufficient internal processes to ensure compliance with § 50.59 thresholds in §§ 50.59 and 53.1550 and related documentation requirements.

C. Revisions to § 52.145, § 53.1521, and § 53.1545

This proposed rule would revise §§ 52.145, 53.1521, and 53.1545, “Updating Final Safety Analysis Reports,” and add § 53.1521(f) and (g), to establish a formal process for amending an SDA and submitting updates to the FSAR. Under this process, any material change to an approved SDA that exceeds the thresholds defined in § 50.59, § 53.1521, or § 53.1550(a)(2) would require prior NRC approval through a structured amendment application.

Additionally, a provision would be added to ensure that the NRC is informed when FSAR changes occur. The SDA holders would be required to submit an updated FSAR within no more than 24 months from the change, if changes have occurred since the last update, ensuring the design documentation remains current and accurate.

While incorporating SDAs into the 10 CFR parts 50 and 53 change processes in §§ 50.59 and 53.1521, respectively, provide flexibility for minor updates, a clearly defined amendment pathway is essential for changes that could affect safety or alter regulatory conclusions. The proposed approach would ensure continued NRC oversight of significant design modifications. The requirement for submitting FSAR changes would be minimized by requiring a submittal only when a change has been made and would provide flexibility to submit the change during the time period within 24 months following the change.

SDA holders would benefit from a well-defined regulatory mechanism to implement necessary design updates. The amendment process would include comprehensive NRC safety evaluation.

D. Conforming Changes to §§ 52.145, 53.1221, and 50.71

To support the changes described in sections A, B, and C under “Parts 50, 52, and 53—Standard Design Approvals,” of this document, the NRC proposes conforming revisions to two key regulatory sections: §§ 52.145 and 53.1221. These regulations would be revised to clarify the finality of SDA approvals, ensuring that once approved, the design is not subject to further NRC review, with exceptions; and address how the NRC may request additional information from SDA holders, particularly to confirm that the design remains technically current and consistent with applicable regulations.

The regulations in § 50.71would be updated to require SDA holders to maintain accurate records of the approved design and any changes made under § 50.59 or § 53.1521, and ensure reporting obligations are consistent with those for other 10 CFR part 52 approvals, such as design certifications and combined licenses.

These updates would ensure regulatory consistency and transparency ( printed page 60738) across NRC licensing processes. They also would reinforce the accountability of SDA holders in maintaining the integrity and traceability of their approved designs.

SDA holders would be subject to similar recordkeeping and reporting standards as other licensees and applicants under 10 CFR parts 52 and 53. This would support NRC oversight and facilitate future licensing actions that reference the SDA.

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

The current regulations in §§ 52.18 and 53.1149(a), both titled, “Standards for review of applications,” require that the NRC prepare an environmental impact statement as part of its review of an application for an early site permit. The NRC proposes to remove this requirement, thereby providing the NRC with flexibility to appropriately establish the scope of the environmental review in accordance with the applicable provisions of the NRC's environmental protection regulations in 10 CFR part 51 and recent amendments to NEPA. This proposed change does not necessarily mean that the NRC would not prepare an environmental impact statement for an early site permit; rather, the NRC would assess on a case-by-case basis the level of environmental review required. This could also include the use of a categorical exclusion or an environmental assessment, which would require significantly less resources and time to prepare, for actions that do not have a reasonably foreseeable significant effect.

10 CFR Part 50—Post-TMI Requirements

The NRC is proposing to amend § 50.34(f) to require 10 CFR part 50 applicants to provide information related to addressing lessons learned from the TMI accident in the same manner as currently required for 10 CFR part 52 applicants. The NRC proposes to delete the applicability statement in § 50.34(f) because it is obsolete. Currently, § 50.34(f) states that the regulations therein are only applicable to 10 CFR part 52 applicants and a list of 10 CFR part 50 applicants that at the time of promulgation were applicants but now no longer are. Section 50.34(f)(2)(i) is redundant to simulator fidelity requirements in § 55.46, and therefore, the NRC proposes to delete it. Section 50.34(f)(2)(vi) is redundant to requirements in § 50.46a, “Acceptance criteria for reactor coolant system venting systems,” and would be deleted.

Next, § 50.34(f)(2)(xxv) is redundant to requirements in appendix E, “Emergency Planning and Preparedness for Production and Utilization Facilities,” to 10 CFR part 50 and would be deleted. The NRC proposes to delete § 50.34(f)(2)(xxiv) because it is redundant to 10 CFR part 50, appendix E, section VI, “Emergency Response Data System.” The regulations in § 50.34(f)(1)(xii), (2)(ix), and (3)(v) are redundant to requirements in § 50.44, “Combustible gas control for nuclear power reactors,” so the NRC proposes to delete them. Section 50.34(f)(3)(vi) requires, for plants that are designed with external hydrogen recombiners, redundant dedicated containment penetrations so that, assuming a single failure, the recombiner systems can be connected to the containment atmosphere. In the “Combustible Gas Control in Containment” final rule (68 FR 54123; September 16, 2003), the NRC eliminated requirements in § 50.44(b)(3) and (c)(3)(ii) for hydrogen recombiners to address design basis accidents. Thus, § 50.34(f)(3)(vi) would also be deleted. Section 50.34(f)(2)(iv) would be edited to delete “console,” to increase flexibility for how safety parameters are to be displayed.

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

A. Minimum Number of Manipulations

The NRC is proposing to revise § 55.31(a)(5) to add a provision that, in lieu of the five significant control manipulations an operator license applicant must perform, the facility licensee may determine the minimum number of required control manipulations in accordance with its Commission-approved training program. The current requirement for five significant control manipulations for reactor operator and senior operator applicants is well-established for the existing fleet of power reactors and NPUFs. The NRC also does not have evidence that the current requirements are unnecessary for existing power reactors and NPUFs. However, the NRC expects that five control manipulations may be unnecessary for certain new designs in the future. In that case, licensees could establish a method and/or criteria for determining the appropriate minimum number of manipulations for operator license applicants to perform to gain experience as part of developing their training programs for operator license applicants.

Additionally, the proposed change would allow for additional evolutions performed on the simulation facility to be credited for the minimum required control manipulations. Currently, the regulation specifies the types of plant evolutions that can “count” for the minimum required control manipulations. These evolutions are applicable to light-water reactors. Section 55.31(a)(5) would be revised to allow for other plant evolutions that affect power or reactivity that are not currently specified there to be credited, which would improve flexibility for plant designs other than light-water reactors.

B. Alternative Examination Standards

The NRC would revise paragraphs (a) and (d) of § 55.40, “Implementation,” to allow the use of criteria other than the criteria in NUREG-1021, “Operator Licensing Examination Standards for Power Reactors,” and NUREG-1478, “Operator Licensing Examiner Standards for Research and Test Reactors,” for examination standards to avoid the NRC having to revise the NUREGs for each new plant design and/or licensees having to submit exemption requests that seek to use a document other than these NUREGs. The NRC expects that different plant designs will result in different tasks for operators at those plants, so that the examinations will be different in scope and content than those for current power reactors and NPUFs. The NRC would still review the structure and content for these examinations, but the proposed rule change would prevent having to have an approved exemption to not use these NUREGs—as currently required by § 55.40.

C. Written Examinations and Operating Tests

The NRC expects that different plant designs ( i.e., plants that are not water-cooled reactors) will result in different tasks and duties for the operators at those plants, so that the content of the examinations may be different in scope and content than those for current water-cooled power reactors and NPUFs. Sections 55.41, 55.43, and 55.45 would be amended to allow for other safety-significant topics to be included in the sample for these examinations.

D. Simplification of Simulator Requirements

The NRC proposes to remove the terms “plant-referenced simulator” and “Commission-approved simulator” from 10 CFR part 55 to simplify and clarify requirements that a simulation facility must meet to be used for operating tests, requalification, and/or meeting experience requirements. As long as a simulation facility meets these requirements, it may be used for those purposes; there would be no need to ( printed page 60739) determine whether the simulation facility is a plant-referenced simulator or if it instead must be approved as a simulation facility by the Commission. The NRC would confirm that a simulation facility complies with regulatory requirements by using existing inspection procedures.

The definition of “simulation facility” in § 53.715(c) would also be amended to add the word “simulator” ( i.e., “simulation facility/simulator”) consistent with the proposed changes to 10 CFR part 55, as both terms are used interchangeably.

E. Requalification Examination and Program Periodicity Requirements

The current requirements for licensed operators and facility licensees for requalification are well-established for the existing fleet of power reactors and NPUFs. However, the NRC anticipates that certain requalification program requirements may be unnecessary for certain new designs in the future. In that case, facility licensees could determine the appropriate requirements for requalification programs, including the minimum requalification examination requirements, using the same systems approach to training process that is currently required for Commission-approved requalification programs.

The NRC proposes to amend § 55.59 by deleting paragraphs (a) and (b) and changing paragraph (c) to be paragraphs (a), (b), and (c), so that § 55.59 only contains the program requirements. Section 55.53(h) would be amended to add a new paragraph (h)(1) that would provide the current § 55.53(h) statement and include the word “successfully” between “shall” and “complete.” Proposed § 55.53(h) would also include new paragraphs (h)(2) and (h)(3) to replace §§ 55.59(b) and 55.59(a)(2)(iii). Sections 55.59(a)(2)(i) and (ii) would be moved to new §§ 55.59(c)(4)(i)(A) and (B). Section 55.59(c)(5) would be amended so that records would be maintained until the end of the subsequent requalification program rather than until a license is renewed because the NRC is proposing to eliminate license renewal.

Additionally, § 53.780(c)(1)(i) would be amended by moving the statement requiring a 24-month duration to (ii) and adding “or other duration as approved by the Commission.” Section 53.780(c)(2)(ii)(C) would be amended by adding “or other periodicity as approved by the Commission” as an alternative to the requirement for a periodicity for a complete requalification program examination. Section 53.780(c)(3) would be amended to say that records must be retained until the end of the subsequent requalification period rather than until a license is renewed because the NRC is proposing to eliminate license renewal. Section 53.730(g)(3) would be amended to require the operator licensing requalification programs required under § 53.780(c) or § 53.815(b) to be in effect within the same timeframes specified in § 50.54(i-1).

F. Requirements for Licensed Medical Examiners

The NRC proposes to amend § 55.21 and § 55.23 to say “a physician or licensed medical examiner” wherever “physician” currently is specified and required. The addition of “or licensed medical examiner” would allow medical professionals, other than physicians, licensed by their state to perform medical exams to conduct the required operator exams. Additionally, a new definition for a “licensed medical examiner” would be added to § 55.4.

Part 53 does not define “physician,” but states that the NRC Form 396 must certify that a physician has conducted the medical examination of the applicant/operator. The NRC Form 396 currently is configured for only an MD or DO to make the certification. Similar changes would be made to § 53.765 and § 53.775 to allow the same flexibilities, with the same definitions added to § 53.020 for clarity and consistency between the two regulatory frameworks.

G. Changes to Treatment of Medical License Conditions

The NRC is proposing several changes in 10 CFR parts 53 and 55 to its medical license conditions requirements that would reduce administrative burden on licensees, facility licensees, and the NRC. These reductions would be accomplished by the following:

a. Creating a new license condition in proposed § 55.53(m) to cover and replace a subset of specific medically related license conditions. Specific license conditions would still be used for other medically related license conditions that would not be in the subset covered by proposed § 55.53(m). The subset that would be covered by proposed § 55.53(m) would be identified on the NRC Form 396.

b. Distinguishing between temporary and permanent medical conditions and clarifying the requirements for how a facility licensee must address temporary conditions to improve clarity and reduce unnecessary correspondence.

H. Changes to Operator License Reporting Requirements

The NRC would delete § 50.74 from 10 CFR part 50. Paragraphs (a) and (b) of current § 50.74 would be moved to § 55.55, § 50.74(c) would be deleted because it is redundant to § 55.25, and § 55.25 would be amended to replace the reference to § 50.74 with a reference to § 55.5.

Paragraph 53.726(b)(3) would be deleted because it is redundant to § 53.770.

Sections 55.53(g) and 55.59(b) would be amended by adding “via the facility licensee.” Section 55.59(a) and (b) would also be moved to § 55.53(h). Conforming changes for § 55.53(g) would be made to §§ 53.785(h) and 53.810(f).

I. Revision to Training Requirements in § 50.120

The NRC would delete the 18 months prior to fuel load requirement from § 50.120(b)(1), add the phrase “as applicable” to the § 50.120(b)(2) list of categories of plant personnel, and add to § 50.120(b)(3) the following statement that would replace the 18 month timeframe: “The requirement in paragraph (b)(1) of this section must be met with sufficient time to ensure that qualified personnel are available when needed.” These proposed changes would improve clarity by stating more clearly the requirement for qualified personnel to be available when they are needed as opposed to an arbitrary date prior to fuel load, which can shift over time, and to improve efficiency by eliminating the need for future exemptions in cases where certain categories of plant personnel are determined not to be necessary at a plant. The proposed changes would affect all 10 CFR part 50 OL or 10 CFR part 52 COL applicants who submit applications or are granted licenses after implementation of the final rule, if this proposed change is made effective in a final rule.

Section 53.830(b) would also be revised to delete “Prior to initial fuel load (or, for a fueled manufactured reactor, prior to initiating the removal of the features to prevent criticality required under § 53.620(d)(1),” and would instead state, “Each holder of an OL or COL under this part must, with sufficient time to provide trained and qualified personnel to operate the facility, establish, implement, and maintain a training program that demonstrates compliance with the requirements of paragraphs (c) and (d) of this section.”

J. Elimination of Wait Times for Re-applications in § 55.35

The NRC proposes to amend § 55.35(a) to allow operators' license applicants to file re-applications ( printed page 60740) regardless of the time elapsed since the previous license denial. The current wait times for submitting applications following a denial are unnecessary as long as the application contains the information already identified in § 55.35(a): the extent of the applicant's additional training since the denial and certification that the applicant is ready for re-examination. Applicants may be ready for re-examination sooner than the wait times specified currently in § 55.35.

K. Operating Tests

The NRC is proposing to amend § 55.40(a) through (d) to allow facility licensees the option of proctoring/administering and grading the operating tests. The NRC would oversee, to an appropriate extent, the facility licensee's proctoring and grading of the operating tests through the inspection program.

L. Plant Walkthrough Requirements in § 55.45

The NRC is proposing to amend § 55.45(b) to combine the existing requirements in § 55.45(b)(1) through (3) in § 55.45(b)(1) and to add the following statement as a new § 55.45(b)(2): “If a facility is under construction, suitable alternatives may be used in lieu of the plant walkthrough portion of the operating test.” NUREG-1021, Section ES-3.7, “Alternatives for In-Plant Job Performance Measures at Plants under Construction,” currently provides guidance that would be applicable to this proposed change and is based on experience using “suitable alternatives” to the plant walkthrough portion of the operating test at VEGP Unit 3. Although the proposed amendment to § 55.45(b) is based on experience licensing operators at a new power reactor that is under construction, the proposed amendment to § 55.45(b) would apply also to non-power facilities that are under construction because similar challenges may exist for future non-power facilities that are under construction.

M. Examination Waiver Requirements

Based on recent experience and knowledge of industry operator staffing trends, the NRC proposes to revise the criteria for waiving any or all of the requirements for a written examination and operating test in § 55.47 in the circumstances where a new unit of standard or modular design is built at a site where one or more of the same kind of units are already operating. There are multiple factors that might justify a waiver that are not identified in the current regulation. These edits would provide greater flexibility and avoid the need for exemptions Applicants could request a waiver and provide supporting justification, and the NRC could then decide whether to approve the waiver on a case-by-case basis. If the proposed changes are made in a final rule, the NRC would place the current § 55.47 criteria in NRC guidance as one way of justifying a waiver request.

The NRC proposes changes to amend § 55.47 by:

a. Revising paragraph (a)(1);

b. Moving the content of paragraphs (a)(2) and (3) to (a)(1);

c. Redesignating paragraphs (b) and (c) as paragraphs (a)(2) and (3), respectively; and

d. Adding a new paragraph (b) for criteria for standard designs.

Section 53.780(f) would be amended to add the similar provisions for waivers for standard designs.

N. Removing Obsolete Information From § 55.5

The NRC would remove information from § 55.5 that will soon be obsolete or is obsolete, thus improving clarity in the regulations. The specific proposed change would delete reference to “electronic information exchange,” or EIE, which the NRC anticipates will be replaced soon.

O. Elimination of License Renewal for 10 CFR Part 55 Operators' Licenses

To remove license terms and license renewal for operators' licenses, the NRC would make the following changes:

a. Delete § 55.57, “Renewals”;

b. Modify § 55.55(a) to delete “(a)” and the statement that the license expires 6 years after the date of issuance, add requirements from § 50.74 to § 55.55 (to report to/notify the Commission within 30 days of termination of employment of a license holder or when license is no longer needed), and to add a statement that a license expires if the Commission or facility licensee determines that the applicant cannot meet the minimum standards for health;

c. Delete § 55.55(b);

d. Delete “53.795” from § 53.120(c)(1) and (2);

e. Delete “and renewal” from the title of § 53.795 so that it reads, “Expiration of operator and senior operator license”;

f. In § 53.795, delete “(a) Expiration. (1)”; the statement in paragraph (1) that the license expires 6 years after the date of issuance; paragraph (2); and paragraph (b); and

g. Conforming changes to § 55.27, § 55.5, and § 55.8 ( i.e., to delete “renewal” and references to § 55.57).

Since all licenses issued prior to implementation of these proposed changes state that the license will expire 6 years from the effective date of the license unless it is terminated, renewed, or upgraded sooner, licensees would need to request a license amendment to remove that statement prior to the expiration date of the license if they wish to prevent it from expiring on the expiration date. Following receipt of a request for a license amendment to remove the statement regarding expiration of the license 6 years from the effective date, NRC would issue a license without that statement. The changes would remove administrative burdens associated with applications for license renewals, which are less of an administrative burden on facility licensees, operators, and the NRC compared to a one-time request for a license amendment to remove the 6-year expiration term.

P. Generally Licensed Reactor Operators for Self-Reliant Mitigation Facilities

The NRC would add a new subpart I to 10 CFR part 55, with new §§ 55.74 through 55.79, specifically for optional use by facility licensees of SRMFs. The proposed subpart I would be used in lieu of other requirements in 10 CFR part 55. The change would reduce regulatory burden for facility licensees, operators at SRMFs, and the NRC by providing an operator licensing process that is appropriately “right sized” for the reduced role of operators in the safe operation of the facility. Instead of requesting multiple exemptions from 10 CFR parts 50 and 55 licensed operator requirements, the facility licensee could instead provide evidence to show it is a SRMF, and then it could use the provisions in proposed subpart I for a general license for reactor operators.

Because 10 CFR parts 50 and 52 contain staffing requirements for licensed operators and senior operators, these requirements would be amended to be performance-based instead of prescriptive. The proposed language is similar to that in 10 CFR part 53 for staffing requirements. Also, definitions for a GLRO and SRMF would be added to § 50.2. The proposed SRMF definition would state that the Commission would determine a facility is a SRMF if it meets the criteria in §§ 53.800(a)(1)-(5) or their equivalent. External guidance for the equivalent criteria could be developed.

Also, minor editorial conforming changes would be made to 10 CFR part 53, subpart F, § 53.800 through 53.820.

10 CFR Part 75—IAEA Notifications

The NRC would delete § 75.13(c) and retain § 75.13(b)(3). The purpose of the revision is to eliminate redundant ( printed page 60741) regulatory language and improve regulatory clarity.

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

The NRC proposes to amend its regulations in 10 CFR parts 50, 52, 70, and 72, “Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste,” to eliminate or modify administrative requirements. This proposal builds on recent rulemaking activities, such as RROAR and the development of the proposed changes to § 50.72 in this proposed rule. The proposed revisions also include additional modifications not considered in other rulemaking efforts as noted.

The proposed modifications would eliminate or modify administrative requirements to reduce unnecessary regulatory burden. The elimination of these requirements would allow the NRC and regulated entities to use available technologies to facilitate necessary exchanges of information. The proposed changes would not alter any requirements related to design or operation of facilities or systems and therefore are of low or no safety significance. Moreover, the proposed revisions would align the 10 CFR parts 50 and 52 requirements with 10 CFR part 53 requirements.

Part 70 provides requirements for submitting effluent reports on a semi-annual basis for each licensee authorized to possess and use special nuclear material for processing and fuel fabrication, scrap recovery, conversion of uranium hexafluoride, or in a uranium enrichment facility. The NRC is proposing a change to § 70.59, “Effluent monitoring reporting requirements,” to eliminate the regulatory requirement for semi-annual reporting for licensee effluent monitoring programs for processing and fuel fabrication, scrap recovery, conversion of uranium hexafluoride, or in a uranium enrichment facility. This information would be collected and retained by the licensee and subject to inspection. Effluent reporting would only be required if the annual radiation doses to the public, resulting from the release result in an estimated total effective dose equivalent, or total effective dose, as applicable, of more than 25 mrem per year, or if quantities of radioactive materials released during the reporting periods are significantly above the licensee's design objectives previously reviewed as part of the licensing action.

However, the NRC is also proposing to add an additional section to § 70.59, noting that the effluent reports would still be required to be submitted for spent fuel reprocessing facilities. Due to the higher safety risk significance of these facilities, annual reporting would still be required.

The proposed changes to § 50.36, “Technical specifications,” would remove notification to the NRC for certain technical specification requirements in paragraphs (c)(1) and (c)(2). The rule also would be modified to remove how the required notifications are to be made in paragraphs (c)(7) and (c)(8). The changes would remove redundant requirements in §§ 50.36, 50.72, and 50.73 that were added to § 50.36 in the revised rule published in the Federal Register on May 27, 1988 (53 FR 19241). That rule revision added the references to §§ 50.72 and 50.73 to provide different reporting requirements for power and non-power reactors. This proposed rule would delete the requirements for non-power reactors consistent with the proposed changes to § 50.72 in this proposed rule based on eliminating low-safety significant nonemergency reporting requirements. Non-power reactor non-compliances covered by the notification requirements currently in § 50.36 are unlikely to be of safety significance. Therefore, the rule revision published in May 27, 1988, that added reference to §§ 50.72 and 50.73 is no longer needed. Notification of certain § 50.36 noncompliances could still be required by the proposed § 50.72, depending on whether criteria in § 50.72 are applicable.

Parts 50 and 53 provide requirements for modifying a licensing basis and when these changes require prior NRC approval. Currently, many of these regulations contain a requirement to submit a report of executed changes that a licensee determined did not require prior NRC approval. This proposed rule would remove these reporting requirements. The licensee's determination that these changes do not require prior approval has been, and will continue to be, subject to NRC oversight through the ROP. The relevant proposed changes would delete the following requirements:

a. Section 50.54(a)(3), which requires that changes to the quality assurance program description that the licensee determines do not reduce the commitments be submitted to the NRC in accordance with the requirements of § 50.71(e).

b. Section 50.55(f)(3) and (f)(4)(i), which require that each construction permit holder and early site permit or manufacturing license holder under 10 CFR part 52 submit changes to the quality assurance program description that does not reduce the effectiveness within 90 days.

c. Section 50.54(q)(5), which requires that a report of changes to the emergency plan made without prior NRC approval be submitted in accordance with § 50.4, “Written communications,” within 30 days after the change is put in effect.

d. Section 50.59(d)(2), which requires the submittal as specified in § 50.4 or § 52.3, as applicable, of a report of any changes, tests, and experiments, including a summary of the evaluation of each made that a licensee determined did not require a license amendment pursuant to § 50.59(c)(2). The report must be submitted at intervals not to exceed 24 months. For combined licenses, the report must be submitted at intervals not to exceed 6 months during the period from the date of application for a combined license to the date the Commission makes its findings under § 52.103(g).

e. Section 53.1550(c)(2), which requires the licensee to submit, as specified in § 53.040, “Written communications,” a report containing a brief description of any departures and changes, including a summary of the evaluation of each. A report must be submitted at intervals not to exceed 24 months. For COLs, the report must be submitted at intervals not to exceed 6 months during the period from the date of application for a COL to the date the Commission makes its findings under § 53.1452(g).

f. Section 53.1565(d)(1)(i), which requires changes to the quality assurance program description that the licensee determines do not reduce the commitments, be submitted to the NRC in accordance with the requirements of § 53.1545. Additionally, § 53.1565(d)(2) which requires that quality assurance program description that the licensee determines do not reduce the commitments for siting, construction, and manufacturing must be submitted to NRC within 90 days.

g. § 53.1565(d)(3)(v), which requires that changes to the emergency plan made without prior NRC approval be submitted within 30 days after the change is put in effect.

Parts 50 and 53 contain requirements to periodically submit to the NRC information related to financial status and insurance or financial security. This proposed rule would eliminate the submittal of this information. Licensees would still be required to comply with the underlying requirements. The ( printed page 60742) relevant proposed changes are to delete the following requirements:

a. Section 50.54(w)(3), which requires that licensees report to the NRC on April 1 of each year the current levels of insurance or financial security they maintain and the sources of this insurance or financial security.

b. Section 50.71(b), which requires each licensee and each holder of a construction permit to submit its annual financial report.

c. Section 53.1720(c), which requires that licensees report to the NRC on April 1 of each year the current levels of insurance or financial security they maintain and the sources of this insurance or financial security.

Section 50.55a contains several requirements to submit reports when portions of the American Society of Mechanical Engineers (ASME) Code are exercised as well as the inservice testing plan each interval. This proposed rule would eliminate the submittal of these reports. The provisions that this proposed rule would eliminate are:

a. Section 50.55a(b)(2)(xliii), which requires licenses to submit the analytical evaluation determining the effects of an out-of-limit condition on the structural integrity of the Reactor Coolant System.

b. Section 50.55a(b)(2)(xlviii), which requires applicants or licensees using the 2021 Edition of Section XI of the ASME Code to submit analytical evaluations performed as required by IWB-3132.3 and IWC-3132.3 of Section XI.

c. Section 50.55a(f)(7), which requires Inservice Testing Program Test and Examination Plans (IST Plans) for pumps, valves, and dynamic restraints (snubbers) prepared to meet the requirements of the ASME Operation and Maintenance Code to be submitted to the NRC as specified in § 50.4 within 90 days of their implementation.

Parts 50, 52, and 53 require that licensees periodically provide revisions to the Final Safety Analysis Report (FSAR) for a facility. Licensees may make changes to the FSAR either without prior approval or through the amendment process, depending on the circumstances. Therefore, the FSAR is a document that is being revised throughout the lifetime of a facility. The regulations in §§ 50.71, 52.3, and 53.1545 require licensees to periodically submit the latest version of the FSAR. This proposed rule would eliminate these requirements. These changes would also eliminate certain other submittals, such as reports required by the Technical Specification Bases control program for operating reactors, that are specifically required to be submitted on a frequency consistent with § 50.71(e). If the NRC requires the current FSAR information for a facility for oversight or a licensing decision, it will obtain that information either through the request for additional information process if required during a licensing process, or through alternate methods if required for oversight. The existing § 50.71 requirements for updating FSARs ties updates to planned outages. This requirement assumes that most maintenance at sites are done during planned refueling outages. The NRC has received multiple exemption requests from this requirement stating that plant modifications are no longer primarily completed during outages and this link is no longer required. Therefore, the proposed modification would no longer tie FSAR updates to refueling outages.

Part 70 requires that licensees periodically provide a brief summary of all changes (not requiring pre-approval under § 70.72, “Facility changes and change process”) made to records required by § 70.62(a)(2) ( i.e., process safety information, integrated safety analysis, and management measures) and revised integrated safety analysis summary pages. Licensees may make changes to process safety information, integrated safety analysis, and management measures either without prior approval or through the amendment process. Therefore, these documents are being revised throughout the lifetime of a facility. This proposed rule would eliminate the requirement to submit revised integrated safety analysis summary pages in § 70.72. If the NRC requires the current revised integrated safety analysis summary for a facility for oversight or a licensing decision, it will obtain that information either through the request for additional information process, if required during a licensing process, or through alternate methods if required for oversight.

Several changes have also been proposed to § 70.32, “Conditions of licenses,” to reduce regulatory burden and provide licensees with additional flexibility in reporting timelines, particularly for programmatic changes that do not reduce the effectiveness of safety and security plans. These changes would extend the timeframes for submitting updates to the NRC without prior Commission approval, while maintaining appropriate oversight and ensuring continued protection of public health and safety.

For the safeguards contingency plan, the reporting period for procedural changes that do not reduce plan effectiveness in § 70.32(g) would be extended from 60 days to 4 months. The NRC uses these reports to verify the continued safety and security of the facilities and reviews the changes to make sure the plan still complies with the regulations. The submittal of these reports allows the NRC to more efficiently review the changes made to the plan. However, the NRC is extending the intervals of these reports as this information is mainly used to conduct inspections, and NRC inspectors can obtain the report in a timely manner through other means. The NRC is additionally seeking comments on the deletion of these reporting requirements.

The NRC is proposing changes to emergency plan reporting requirements in § 70.32(i) recognizing that such reports are not required for routine oversight activities. Under current regulations, licensees may make changes to their approved emergency plans without prior Commission approval, provided those changes do not decrease the effectiveness of the plan. These changes must be reported to the NRC within 6 months. The proposed revision would remove the requirement for licensees to submit summaries of emergency plan changes that do not reduce plan effectiveness. This change is intended to reduce administrative burden while maintaining regulatory oversight through inspections. The absence of readily available up-to-date emergency plan updates could also affect NRC response and coordination efficiency with offsite response organizations during an incident. However, the NRC considers the impact on the ability of the agency to effectively respond to a fuel cycle facility event to be minimal, due to the nature of potential fuel cycle facility accidents and the ability to obtain these documents after the event, and the NRC safety mission would therefore be maintained under this proposed change.

The regulations under 10 CFR part 72 have some administrative requirements for independent spent fuel storage installations (ISFSIs) co-located with reactor sites that are similar to those found in 10 CFR part 50. In order to be consistent with changes being made to certain administrative requirements in 10 CFR part 50, changes to some reporting requirements found in 10 CFR part 72 would also be proposed. This proposed rule would modify 10 CFR part 72 to remove the submittal of reports required by § 72.44(f) for emergency plan changes and § 72.70(c)(6) for FSAR updates. This proposed rule would eliminate the requirement to submit reports of changes to this document but instead would require that the reports of ( printed page 60743) changes be maintained as a record. If the NRC should require the most current information on an ISFSI for oversight or licensing purposes, that information would be obtained either through the request for additional information process, if required during a licensing review, or through alternate methods if required for oversight. Conforming changes would also be proposed for § 72.248(c)(1), (2), (3) and (6).

Part 72 contains requirements to periodically submit information related to the financial levels, insurance, and decommissioning funding for ISFSIs. This proposed rule would eliminate the submittal of some of this information. A licensee's continued compliance with the financial requirements as well as insurance and decommissioning funding would continue to be subject to NRC oversight. The relevant proposed changes would be to delete the financial requirements in § 72.80(b), which requires each ISFSI licensee to submit its annual financial report, unless it already submits a Form 10-Q with the Securities and Exchange Commission.

Some of the non-emergency notifications required by § 72.75, “Reporting requirements for specific events and conditions,” specifically in § 72.75 (b), (c) and (d). This proposed rule would eliminate the submittal of some of these notifications. The current reports are seen as duplicative as the information is already provided to NRC inspectors informally and dispositioned accordingly. The relevant proposed change is to delete the following non-emergency notification requirements (these are conforming changes to changes proposed to § 50.72):

• Section 72.75(b)(2), which requires 4-hour reporting of an event or situation related to the health and safety of the public or onsite personnel to the NRC, if a news release or reporting to another Government agency has been made or is planned.
• Section 72.75(c)(3), which requires 8-hour reporting of any event requiring the transport of a radioactively contaminated individual to an offsite medical facility for treatment.

Next, §§ 50.4 and 53.040 describe how certain information is to be submitted to the NRC. To conform with the deletion of reporting requirements in 10 CFR parts 50 and 53, this proposed rule would modify §§ 50.4 and 53.040 to remove specific requirements for submitting reports required by:

a. Sections 50.54(a)(3), 50.55(f)(3), and 53.1565 related to quality assurance plan changes.

b. Sections 50.54(q)(5) and 53.1565 for emergency plan changes.

c. Sections 50.71(e) and 53.1545 for FSAR updates.

d. In addition, this proposed rule would delete or modify requirements to remove reference to obsolete or unnecessarily prescriptive communication methods. These changes would include:

e. Section 50.91(a)(6)(i)(B), which requires the use of local media to provide notice of consideration for an exigent amendment ( i.e., where sufficient time to provide notice via the Federal Register is not available). This proposed rule would eliminate the requirement to use local media. This modification would permit the use of other technologies, such as the NRC website, to provide notice so that notice is more visible to the public.

f. Section 50.91(b)(3) and (b)(4), which require the NRC to telephone the State official for consultation prior to issuance of an amendment. This proposed rule would modify this requirement to say “contact” instead of “telephone,” which would permit the NRC to use other means, such as through email or other technology, to contact the state office before issuance of an amendment.

The proposed revisions to 10 CFR parts 50 and 52 would align these parts with 10 CFR part 53 requirements.

Miscellaneous Corrections

The NRC is proposing the following updates that are corrective or miscellaneous updates:

a. Amending § 50.8, “Information collection requirements: OMB approval,” to include the new OMB Control number for § 50.55a, “Codes and standards,”

b. Changing the term “Website” to “website” in the applicable 10 CFR Parts,

c. Amending weblinks to include “https” in the applicable 10 CFR Parts,

d. In § 50.54, updating an incorrect reference to paragraph (i)(1) to the correct paragraph, (i-1), and

e. Amending § 55.8, “Information collections requirements: OMB approval,” to include § 55.46, “Simulation facilities.”

V. Specific Requests for Comments

The NRC is seeking advice and recommendations from the public on this proposed rule. The NRC is particularly interested in comments and supporting rationale from the public on questions on the following topic areas.

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

Question 1: For existing 10 CFR part 70 licensees, does the new definition for “basic component” add or reduce the number of components that would require commercial grade dedication under the current 10 CFR part 21 rule? Please provide a basis for your response.

Question 2: For new 10 CFR part 70 licensees, does the new definition for “basic component” provide clarity on the extent to which commercial grade dedication activities should be performed? Please provide a basis for your response.

Question 3: For new and existing 10 CFR part 70 licensees, do the rulemaking changes, especially the new definition of “basic component”, add regulatory clarity and reduce burden or introduce unintended consequences such as additional burden and implementation concerns? Can the unintended consequences be mitigated by implementation guidance and outreach efforts? Please provide a basis for your response.

Question 4: The NRC has proposed concurrent rulemakings that could allow for other quality assurance frameworks. Should 10 CFR part 21 be expanded to be more inclusive of those quality assurance frameworks?

Question 5: For new and existing power reactor licensees, do the rulemaking changes, especially the new definition of “dedicating entity”, add regulatory clarity and reduce burden or introduce unintended consequences such as additional burden and implementation concerns? Can the unintended consequences be mitigated by implementation guidance and outreach efforts? Please provide a basis for your response.

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

Question 6: Does any portion of the proposed update to NUREG-1022, Revision 4, warrant additional clarification to better inform the level of detail expected in nonemergency event reports, to help minimize the burden on reactor operations during event response? Particularly, should any of the discussion in sections 2, 3, or 4, or in section 5 for Licensee Event Reports, be expanded, clarified, updated, or otherwise modified? Please provide a basis for your response.

Question 7: Is any additional guidance needed to minimize uncertainty or promote consistency in reporting, or to address situations that raise doubt as to whether immediate reporting is necessary? This includes situations involving invalid, inadvertent, or anticipated safety system actuations during maintenance activities, particularly those associated ( printed page 60744) with the reactor protection or emergency core cooling systems. Please provide a basis for your response.

Question 8: Does any portion of the proposed update to NUREG-1022, Revision 4, warrant additional guidance to improve the common understanding of the need for NRC situational awareness and commensurate response to various plant events and conditions or the level of detail required in the reports to support that need? Please provide a basis for your response.

Question 9: Does the proposed rule or draft guidance create conditions that would be contrary to the rule's purpose and objectives? If so, which portions are contrary to the rule's purpose and objectives and why? Are there unintended consequences? If so, what are the unintended consequences, and how should they be addressed? Please provide a basis for your response.

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

Question 10: For facility licensees/industry: Are there any potential challenges with gathering necessary information to submit an LTP within 2 years of permanent cessation of operations? The NRC is considering various approaches to mitigate foreseeable challenges with submitting an upfront LTP, including amendments to existing regulations and guidance. For example, what techniques would licensees need to employ to develop adequate site characterization to support the upfront LTP? Please provide a basis for your response.

10 CFR Parts 50, 52, 53, 70, and 72—Backfitting and Issue Finality Provisions

Question 11: The regulations in 10 CFR parts 52 and 53 allow a manufacturing license to reference a standard design certification. Additionally, a construction permit, operating license, or combined license issued under 10 CFR parts 50, 52, or 53 may authorize the construction or use of a manufactured reactor under such a manufacturing license.

Should the NRC revise its regulations in 10 CFR parts 50, 52, or 53 to clarify the applicability of the backfitting and issue finality provision under these circumstances? Please provide the basis for your response and identify the specific matters that should be addressed under each of the relevant backfitting and issue finality regulations.

10 CFR Parts 52 and 53—Notification of Initial Fuel Load

Question 12: Are there any concerns with the reduced frequency of notifications regarding the schedule for initial loading of fuel or the schedule for initiating the removal of the features to prevent criticality for licensees installing fueled manufactured reactors under a 10 CFR part 53 COL? Please provide a basis for your response.

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

To support the development of the final rule and ensure stakeholder perspectives are considered, the NRC invites comments on the following targeted questions:

Applicability of § 50.59 to SDA Holders

Question 13: Do stakeholders agree that applying the § 50.59 change control process to SDA holders is appropriate and consistent with the intent of the regulation? Please provide a basis for your response.

Question 14: Are there unique challenges or considerations for SDA holders in implementing § 50.59 compared to licensees? Please provide a basis for your response.

Recordkeeping Burden and Clarity

Question 15: Is the proposed recordkeeping requirement for SDA holders under § 50.71 sufficiently clear and achievable? Please provide a basis for your response.

Question 16: Are there alternative approaches to documenting and retaining design changes that would reduce burden while maintaining transparency? Please provide a basis for your response.

Design Finality and Flexibility

Question 17: Should the NRC consider additional criteria or triggers for reevaluating SDA validity over time? Please provide a basis for your response.

Impact on Referencing Applicants

Question 18: How will the proposed changes affect applicants who reference an SDA in a combined license or design certification application? Please provide a basis for your response.

Guidance Needs

Question 19: Are existing NRC guidance documents ( e.g., RG 1.187, RG 1.70) sufficient to support implementation of the proposed changes or are there any specific updates or new guidance that stakeholders would find helpful?

Future Enhancements

Question 20: Should the NRC consider expanding the applicability of § 50.59 to other 10 CFR part 52 licensing instruments ( e.g., manufacturing licenses)? Please provide a basis for your response.

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

Operating Tests

Question 21: Currently, power reactor facility licensees may prepare, proctor and grade the written examinations required by § 55.41 and § 55.43 and may prepare the operating tests required by § 55.45 ( i.e., the dynamic simulator scenarios and the plant walkthrough), subject to certain conditions contained in § 55.40 (b). The NRC proposes to provide power reactor and non-power reactor facility licensees the option of proctoring and grading the operating tests. What benefits and challenges could such flexibility introduce and how could the NRC adapt accordingly? Please provide a basis for your response.

Question 22: For facility licensees/industry: Is the conclusion in the regulatory analysis correct that it would be cost effective for the facility to use examiner staff to administer and grade the operating tests with NRC oversight rather than to have the NRC administer and grade the operating tests, with the costs associated with both options to be reimbursed by the licensee? Specifically, is the conclusion in the regulatory analysis correct that there would be no new costs incurred by facility licensees to utilize the new provisions? Please provide a basis for your response. For facility licensees/industry: Would it be appropriate to conclude that the revised rule would provide additional flexibility for facility licensees with scheduling examinations? Please provide a detailed description of the how the proposed rule would provide additional flexibility.

Plant Walkthrough Requirements in § 55.45

Question 23: The NRC is particularly interested in comments and supporting rationale from the public on the continued relevance of the plant walkthrough as part of the operating test for power reactor applicants and proposed recommendations related to potentially modifying or removing the requirement. If the requirement was modified or deleted for power reactor applicants, the NRC would expect to revise the guidance in NUREG-1021 ( e.g., re-evaluate grading for job performance measures, and potentially ( printed page 60745) increase the number of control room systems and/or administrative job performance measures).

Should the NRC eliminate or amend the plant walkthrough requirement as part of the operating test administered to applicants at power reactors, and if so, what revisions should the NRC make to NUREG-1021? Please provide a basis for your response.

Examination Waiver Requirements

Question 24: As discussed in § 55.47, “Waiver of examination and test requirements,” the NRC may waive any or all of the requirements for a written examination and operating test, if it finds that the applicant has met the criteria in § 55.47(a)(1)-(3). According to § 55.47(a)(1), the NRC must find that the applicant has had extensive actual operating experience at a comparable facility, as determined by the Commission, within 2 years before the date of application. The basis for the original 2-year requirement was that “extended” time ( i.e., greater than 2 years) away from licensed duties would negatively impact operator performance.

The NRC expects that there will be situations where the 2-year requirement may not be met, but alternative justification could be provided. Should the NRC proceed with this approach to the 2 years or consider an alternative? Please provide a basis for your response. If the NRC should consider an alternative, please explain the alternative.

Elimination of License Renewal for 10 CFR Part 55 Operators' Licenses

Question 25: In eliminating renewal, the NRC identified that certain record retention requirements tied to renewal would need to be adjusted. Specifically for § 55.27, medical records would be retained for the entire term of the license for the operator or senior operator rather than just the period of renewal (6 years). It is also the NRC's understanding that facility licensees already retain medical records for the entire term of the license. For facility licensees/industry would there be an increase in burden on record keeping/record retention?

Operator Staffing Requirements

Question 26: The NRC proposes changes to § 50.54(m) to accommodate the addition of new requirements for generally licensed reactor operators in 10 CFR part 55. The proposed changes would provide an option for facility licensees to submit a staffing plan for NRC approval to be used in lieu of the requirements in § 50.54(m). The NRC is seeking input specifically from facility licensees on how a transition would occur to an approved staffing plan.

10 CFR Parts 50, 52, and 53—Record Keeping and Reporting Requirements

Question 27: For industry stakeholders: Are there additional reporting, recordkeeping, or administrative requirements in 10 CFR that the NRC should consider for revision or elimination as part of this rulemaking? If so, please identify the specific requirements, explain why they may no longer be necessary or effective, and describe how their removal or modification would support regulatory efficiency without compromising safety, security, or environmental protection.

Question 28: For industry stakeholders: Are there specific reporting requirements that could be improved through the use of alternative, technology-inclusive submission methods ( e.g., automated systems, standardized digital forms, or secure online portals)? Please identify specific reporting processes that would benefit from such improvements and describe the potential impact on efficiency, accuracy, or burden reduction.

Question 29: For government agencies, academic institutions, and public interest organizations: The proposed rule includes several changes to reporting requirements related to security plan updates, FSAR updates, quality assurance plan updates, and effluent reporting. How would the elimination of certain reporting requirements and reduced accessibility to related records affect situational awareness for State and local officials, public communication, or research efforts? Please describe any anticipated impacts on your ability to monitor, analyze, or respond to nuclear-related events, including potential costs or limitations to public transparency, academic study, or emergency preparedness.

Question 30: For government agencies, public interest organizations, and industry stakeholders: The NRC is proposing to eliminate some reporting requirements but is electing to retain the requirement to submit summaries of security plan changes and changes in the physical protection plan, and safeguards contingency plan for material in transit made without prior NRC approval. The NRC is seeking comments on the implications of preserving or eliminating these requirements. Does the retention of these requirements aid in public transparency, support state and local official situational awareness, and for the purposes of oversight, policy formulation, and/or emergency response sufficiently justify the associated cost burden on licensees? In retaining the requirement to report security plan and physical protection plan and, for material in transit, safeguard contingency plan, the NRC considered the potential savings of a change to the requirement, balanced with a potential increase in inspection resources. For example, if reporting were no longer required, inspectors would be required to request the current plans as a part of inspection preparation from licensees prior to conducting the inspection. The NRC considered this to be an overall increase in NRC and licensee resources. Please provide feedback if you disagree with this assessment.

Question 31: For government agencies, public interest organizations, and industry stakeholders: The NRC is proposing to eliminate the requirement for licensees to submit reports of emergency plan changes made without prior NRC approval within 30 days of implementation. The NRC is seeking input on whether the removal of this timely reporting and documentation requirement could affect the ability of offsite response organizations, local governments, and the public to remain informed about modifications that may influence emergency response coordination or protective action strategies. In considering this change, the NRC acknowledges that, in the event of an incident, the absence of readily accessible records of emergency plan changes could hinder communication and delays response efforts. However, due to the unlikely event of an emergency, and the ability to readily receive a copy of the current report, the impact would be minimal. Stakeholders are encouraged to provide feedback on the potential impacts of this proposed change including any concerns related to transparency, situational awareness, and emergency preparedness effectiveness.

Question 32: The proposed rule includes several changes to reporting requirements. Examples of such changes include the elimination of the requirements for periodic submission of Updated Final Safety Analysis Report (UFSAR) and licensee-controlled changes to their security plan. These changes are supported by the experience of the operating fleet in exercising the programs that allow such licensee-controlled changes. Advanced reactor licensees would not only lack this experience but are also likely to encounter more frequent updates early in their operating history. Further, certain advanced reactor licensees may not have resident inspectors resulting in potential challenges to the timely availability of key documents ( e.g., the ( printed page 60746) UFSAR), especially for mission-critical NRC functions such as incident response. What guardrails or information sharing approaches should the NRC consider to mitigate such situations?

VI. Regulatory Flexibility Certification

As required by the Regulatory Flexibility Act of 1980, 5 U.S.C. 605(b), the Commission certifies that this rule, if adopted, will not have a significant economic impact on a substantial number of small entities. This proposed rule affects only the licensing and operation of nuclear power plants and fuel facilities. The companies that own these plants do not fall within the scope of the definition of “small entities” set forth in the Regulatory Flexibility Act or the size standards established by the NRC (10 CFR 2.810).

VII. Regulatory Analysis

The NRC has prepared a draft regulatory analysis on this proposed regulation. The analysis examines the costs and benefits of the alternatives considered by the NRC. The NRC requests public comment on the draft regulatory analysis. The regulatory analysis is available as indicated in the “Availability of Documents” section of this document. Comments on the draft analysis may be submitted to the NRC as indicated under the ADDRESSES caption of this document.

VIII. Backfitting and Issue Finality

This section describes the backfitting and issue finality implications of this proposed rule and the draft guidance documents described in Section XV, “Availability of Guidance,” of this document, as applied to applicants and holders of pertinent NRC approvals. The NRC's backfitting and issue finality provisions are provided in 10 CFR parts 50, 52, 53, 70, 72, and 76. MD 8.4 describes the Commission's policies on backfitting and issue finality.

The backfitting provisions relevant to this proposed rule are in §§ 50.109, 53.1390, 70.76, and 72.62. Commercial NPUF licenses issued under § 50.22, nuclear power reactor licenses issued under 10 CFR parts 50, 52, and 54, and certain related NRC approvals issued under 10 CFR parts 50 and 52 are within the scope of § 50.109. Special nuclear materials licenses issued under 10 CFR part 70, subpart H, and ISFSI licenses under 10 CFR part 72 are within the scope of §§ 70.76 and 72.62, respectively. Monitored retrievable storage installations are also within the scope of § 72.62, but no such facilities have been licensed.

There are various issue finality regulations for nuclear power reactor-related approvals under 10 CFR part 52. The issue finality provisions for early site permits and design certifications in §§ 52.39 and 52.63, respectively, are relevant to this proposed rule. The issue finality provisions for combined licenses in § 52.98 and in the design certification rules in the appendices to 10 CFR part 52 are also relevant, in that they identify which matters are within the scope of §§ 50.109 or 52.63 for combined licenses that reference a design certification rule. The issue finality provisions related to the renewal of an early site permit or design certification in §§ 52.31 and 52.59, respectively, only apply to the NRC's review of an application for renewal and not to the proposed rule changes for this rulemaking. Similarly, the issue finality provisions for SDAs in § 52.145 only apply to the NRC's review of an application referencing the SDA and not to the proposed rule changes for this rulemaking.

This proposed rule would not result in backfitting as defined in §§ 50.109, 70.76, and 72.62 and would not affect issue finality as provided for in 10 CFR part 52. This proposed rule would not affect the issue finality of manufacturing licenses issued under 10 CFR part 52 because no manufacturing licenses are in effect. This proposed rule would also not result in backfitting or affect issue finality of an approval under 10 CFR part 53 because no NRC approvals have been issued under 10 CFR part 53. The backfitting provisions in § 76.76, “Backfitting,” are also not relevant to this proposed rule because there are no certificates of compliance for gaseous diffusion plants in effect. Therefore, the backfitting and issue finality implications for 10 CFR parts 53 and 76 and for manufacturing licenses under 10 CFR part 52 are not considered further in this assessment.

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

The proposed changes to § 50.55(e) would eliminate duplicative requirements and relocate two requirements to 10 CFR part 21. Eliminating duplicative requirements would not result in a modification of or addition to systems, structures, components, or design of a facility; or the design approval or manufacturing license for a facility; or the procedures or organization required to design, construct, or operate a facility. The requirements proposed to be relocated to 10 CFR part 21 would be applicable to the same entities that were subject to those requirements when they were in § 50.55(e). Consequently, the relocated requirements would not result in a modification of or addition to systems, structures, components, or design of a facility; or the design approval or manufacturing license for a facility; or the procedures or organization required to design, construct, or operate a facility. Because these proposed changes would not meet the definition of backfitting as defined in §§ 50.109 and 72.62, the proposed changes would not constitute backfitting and would not affect issue finality under 10 CFR part 52.

The proposed changes to the definitions of “commercial grade item,” “dedication,” and “dedicating entity” in § 21.3 would not impact any entity that has implemented a commercial grade dedication process. Therefore, these proposed changes would not constitute backfitting as defined in §§ 50.109, 70.76, and 72.62, and would not affect issue finality under 10 CFR part 52.

The proposed revision of the definition of “basic component” in § 21.3 would provide clarity for fuel fabrication, fuel conversion, and plutonium processing and fuel fabrication plants and spent fuel processing facilities that are required to develop an integrated safety analysis in accordance with 10 CFR part 70, subpart H. The proposed changes to the definition of “basic component” would reduce the number of components that would be managed under a commercial grade dedication program for such facilities. Components that would no longer meet the definition of a basic component, because they are at a lower-risk level, would not be required to be removed from the licensee's commercial grade dedication program. Therefore, the proposed change would be a non-mandatory relaxation of requirements for fuel facilities licensed under 10 CFR part 70, subpart H. As explained in MD 8.4, non-mandatory relaxations of regulations generally do not meet the definition of “backfitting.” Thus, the proposed change would not meet the definition of backfitting in § 70.76. There are no backfitting or issue finality implications for fuel facilities licensed under only 10 CFR part 40 because 10 CFR part 40 licenses are not within the scope of any backfitting or issue finality provisions.

The proposed clarification that commercial grade dedication is not applicable to nuclear material users would not have any backfitting or issue finality implications because the licenses for the affected nuclear materials users are not within the scope of any backfitting or issue finality provisions. ( printed page 60747)

The proposed change to the NRC's preferred method of initial notification of a defect or failure to comply under § 21.21(d) from facsimile to email would provide updated information on how to contact the NRC Operations Center. The current regulations in § 21.21 require entities subject to 10 CFR part 21 to have procedures to identify, evaluate, and report defects and failures to comply. Procedures required by § 21.21 can include procedures used to assess procedures required to design, construct, or operate a facility for potential defects and failures to comply. The proposed continued availability of telephone notification would ensure that licensees can comply with § 21.21(d) without adopting the proposed email method. Licensees could continue using the existing processes and procedures and still be in compliance with this proposed rule. This proposed change to the methods to contact the NRC would not meet the definition of “backfitting” in §§ 50.109, 70.76, and 72.62 because the proposed change would provide a voluntary alternative. Additionally, the requirements of § 21.21 are not site characteristics, design parameters, or terms and conditions on an early site permit and they are not certification information under 10 CFR part 52. Therefore, the proposed change to the methods to contact the NRC would not affect issue finality as provided for in 10 CFR part 52.

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

The NRC is proposing the elimination of certain reporting requirements in § 50.72(b) for nonemergency events at nuclear power plants authorized to operate under 10 CFR parts 50 or 52. The proposed changes would be a nonmandatory relaxation of requirements because affected licensees may continue to voluntarily submit such reports. Therefore, the proposed changes to the reporting requirements would not meet the definition of “backfitting” under § 50.109.

The proposed elimination of certain nonemergency event reporting requirements would not affect the issue finality of a combined license under 10 CFR part 52. The issue finality provisions for combined licenses in § 52.98 provide, in relevant part, that the Commission may not modify, add, or delete any term or condition of a combined license except in accordance with the provisions of § 50.109. As explained previously, eliminating nonemergency reporting requirements would not meet the definition of “backfitting” in § 50.109. Additionally, the regulations in § 50.72 are not within the scope of matters addressed under other 10 CFR part 52 approvals for which issue finality is provided. For this reason, the proposed eliminations also would not affect the issue finality of a combined license or any other approval under 10 CFR part 52.

The NRC also proposes to provide nuclear power reactor licensees with the option to use technology-inclusive alternatives to the ENS for making a nonemergency event notification to the HOC. Because an alternative reporting method would be voluntary, such that licensees would still be required to submit the event reports but would not be required to use an alternative method of submitting the reports, this proposal would not meet the definition of “backfitting” in § 50.109 or affect the issue finality of a combined license.

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

The NRC is proposing several decommissioning licensing enhancements for production and utilization facilities licensed under 10 CFR parts 50, 52, and 53, as applicable. These enhancements would include eliminating or relaxing certain requirements, adding options for licensees, and making clarifications and corrections. The proposed eliminations and relaxations of certain requirements would be either nonmandatory, meaning that licensees would not need to make changes to their current procedures to comply with the revised requirements, or applicable to only the NRC. The proposed provision of voluntary options would not eliminate the current regulatory options and would not be imposed on licensees. The clarifications and corrections would not result in a modification of or addition to systems, structures, components, or design of a facility; or the design approval or manufacturing license for a facility; or the procedures or organization required to design, construct, or operate a facility. One set of proposed changes, concerning the financial qualifications standard, would apply to only applicants. Therefore, the proposed decommissioning licensing enhancements would not meet the definition of “backfitting” in § 50.109 and would not affect issue finality as provided for in 10 CFR part 52.

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

The proposed changes to §§ 50.33, 50.75, 50.82, and 53.1045 would relax certain requirements related to financial qualifications and decommissioning trust fund use under 10 CFR parts 50, 52, or 53 and would make these requirements consistent across all of these parts of the NRC's regulations. These changes would be nonmandatory relaxations because licensees would not need to make changes to their current practices to comply with the proposed requirements. Therefore, these proposed changes would not meet the definition of “backfitting” in § 50.109 and would not affect issue finality as provided for in 10 CFR part 52.

10 CFR Part 50—Risk Informed Seismic Design

The proposed changes to appendix S to 10 CFR part 50 would provide an option for applicants seeking certain approvals under 10 CFR part 50 or 10 CFR part 52 to establish multiple DBGMs based on the risk-significance and safety functions of SSCs. The proposed change would apply to only future applicants and therefore would not constitute backfitting under 10 CFR part 50 or affect the issue finality of a 10 CFR part 52 approval. Applicants and potential applicants (for licenses, permits, and other regulatory approvals) generally are not within the scope of the backfitting or issue finality regulations. Those regulations include language delineating when those provisions begin; in general, the backfitting and issue finality regulations begin upon the issuance of the license, permit, or other approval. Therefore, the proposed changes to appendix S to 10 CFR part 50 would not meet the definition of “backfitting” in § 50.109 and would not affect issue finality as provided for in 10 CFR part 52.

10 CFR Part 50—Probabilistic Risk Assessment Requirements

The proposed changes to § 50.69 would allow holders of an LWR CP under 10 CFR part 50 and holders of an LWR COL or ML and applicants for an LWR DC under 10 CFR part 52 to utilize the provisions of this regulation. Because § 50.69 is a voluntary, alternative regulation and the proposed changes would only extend its applicability, they would not meet the definition of “backfitting” under § 50.109 or affect issue finality as provided in 10 CFR part 52.

10 CFR Parts 50, 52, 53, 70, and 72—Backfitting and Issue Finality Provisions

The proposed amendments to the backfitting and issue finality provisions would change requirements that apply only to the NRC. Therefore, the proposed amendments to the backfitting ( printed page 60748) and issue finality provisions would not meet the definitions of “backfitting” in §§ 50.109, 70.76, and 72.62 or affect issue finality for NRC approvals issued under 10 CFR part 52.

10 CFR Parts 52 and 53—Notification of Initial Fuel Load

The proposed changes would relax the notification frequency for updates to the schedule for initial fuel loads for combined licenses for power reactors issued under 10 CFR part 52 or 10 CFR part 53. The proposed changes would be a nonmandatory relaxation of requirements because affected licensees could continue to submit such notifications according to the current schedule requirements and still satisfy this proposed rule. Therefore, the proposed changes to the reporting requirements would not meet the definition of “backfitting” under § 50.109 or affect issue finality as provided for in 10 CFR part 52.

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

This proposed rule would eliminate requirements for certain applications under 10 CFR part 52 to provide information regarding generic issues and operating experience. The proposed change would apply to only future applicants and would not affect matters for which issue finality is provided for existing approvals under 10 CFR part 52.

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

This proposed rule would eliminate the expiration date for SDAs and provide processes to allow the holder of an SDA to make changes to the SDA. The elimination of the 15-year expiration date for an SDA would not modify existing SDAs as it would apply to the future issuance of SDAs by the NRC. The proposed change control processes would not meet the definition of “backfitting” because they would be part of an NRC-designed administrative change process that would not be required to design, construct, or operate a facility. The SDA holder would not be required to make changes to the design or its procedures as a result of these proposed rule changes unless the SDA holder would voluntarily choose to implement the proposed change control process. Therefore, these proposed rule changes would not meet the definition of “backfitting” under § 50.109 and would not be within the scope of any issue finality provisions in 10 CFR part 52 because such provisions would apply only to an applicant referencing an SDA and not to the SDA holder itself. This proposed rule would also require an SDA holder that makes changes to the design to report and retain records of such changes. These proposed requirements would not meet the definition of backfitting because, even if an SDA holder would need to change a procedure to meet these proposed reporting and recordkeeping requirements, that procedure would not be required to design, construct, or operate a facility. These reports would be needed to ensure continued NRC oversight.

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

The proposed amendments would eliminate a requirement for the NRC to prepare an environmental impact statement as part of its review of an application for an early site permit. Because the proposed changes would apply to the NRC, they would not meet the definition of “backfitting” under § 50.109 or affect issue finality as provided for in 10 CFR part 52.

10 CFR Part 50—Post-TMI Requirements

The post-TMI requirements in § 50.34(f) apply to certain applications for power reactor related approvals. The proposed changes to § 50.34(f) would make § 50.34(f) applicable to applications for power reactor construction permits and operating licenses under 10 CFR part 50 and remove license applications that are no longer relevant from the list. The proposed amendments would also delete redundant requirements in § 50.34(f)(1)(xii), (f)(2)(i), (f)(2)(vi), (f)(2)(ix), (f)(2)(xxiv), (f)(2)(xxv), (f)(3)(v) and (f)(3)(vi), and revise (f)(2)(iv) to increase flexibility. The proposed changes to § 50.34(f) would not meet the definition of “backfitting” in § 50.109 and would not affect issue finality as provided for in 10 CFR part 52 because the proposed changes would apply to only applicants for power reactor construction permits and operating licenses under 10 CFR part 50.

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

Backfitting and issue finality regulations do not appear in 10 CFR part 55, and part 55 power reactor operator licensees are not within the scope of the backfitting provisions in § 50.109 or the finality provisions in 10 CFR part 52. However, the proposed changes to the operator licensing program requirements in this proposed rule could result in modifications to the procedures or organizations required to operate a facility. Therefore, the NRC considered the backfitting and issue finality implications of the proposed amendments to the operator licensing programs for facilities licensed under 10 CFR parts 50 and 52.

A. Minimum Number of Manipulations

The proposed amendments would provide voluntary alternatives for facility licensees to address control manipulations for operator license applicants. Furthermore, the proposed changes for operator license applicants and licensed operator training would not modify procedures or organizations required to operate a facility. Therefore, the proposed changes to the operator license application provisions regarding control manipulations would not meet the definition of “backfitting” in § 50.109 or affect issue finality as provided for in 10 CFR part 52.

B. Alternative Exam Standards

The proposed amendments would allow for alternative exam standards for operators while retaining the current exam standards. These proposed changes would provide voluntary alternatives for facility licensees. Furthermore, changes for operator license applicants and licensed operator training would not involve a procedure or organization required to operate a facility. Therefore, the proposed changes to allow for alternative exam standards would not meet the definition of “backfitting” in § 50.109 or affect issue finality as provided for in 10 CFR part 52.

C. Written Exams and Operating Tests

The proposed amendments would clarify topics on the examination and tests for operator license applicants at non-water-cooled reactors. Currently, no facility licensees of non-water-cooled reactors have developed NRC licensing examinations. The proposed changes would not affect any currently operating or decommissioning reactor facilities. Furthermore, proposed changes to the content of examinations and tests for operator license applicants would not modify procedures or organizations required to operate a facility. Therefore, the proposed changes to the operator licensing written examination and test provisions would not meet the definition of “backfitting” in § 50.109 or affect issue finality as provided for in 10 CFR part 52.

D. Simplification of Simulator Requirements

The proposed amendments would simplify the simulator requirements. The NRC is not proposing any changes to the criteria a simulator must meet to ( printed page 60749) be used for its intended regulatory purposes. The proposed changes for simulators used for training and examining operator license applicants and licensed operator training would not involve a procedure or organization required to operate a facility. Therefore, the proposed changes to simulator requirements would not meet the definition of “backfitting” in § 50.109 or affect issue finality as provided for in 10 CFR part 52.

E. Requalification Examination and Program Periodicity Requirements

The proposed amendments would provide voluntary options for facility licensees and operators to address requalification requirements. Furthermore, the proposed changes for requalification would not modify procedures or organizations required to operate a facility. Therefore, the proposed changes to requalification requirements would not meet the definition of “backfitting” in § 50.109 or affect issue finality as provided for in 10 CFR part 52.

F. Requirements for Licensed Medical Examiners

The proposed amendments would provide nonmandatory relaxations of current requirements regarding licensed medical examiners in 10 CFR part 55. Therefore, these proposed changes would not meet the definition of “backfitting” in § 50.109 or affect issue finality as provided for in 10 CFR part 52.

G. Changes to Treatment of Medical License Conditions

The proposed changes to the treatment of license conditions would not modify procedures or organizations required to operate a facility. Therefore, the proposed changes to the treatment of medical license conditions for operators would not meet the definition of “backfitting” in § 50.109 or affect issue finality as provided for in 10 CFR part 52.

H. Changes to Operator License Reporting Requirements

The proposed amendments would relocate certain operator license reporting requirements and delete redundant reporting requirements. These proposed changes would not result in the underlying requirements changing. The proposed change to § 55.53(g) would not require changes to the procedures or organizations required to operate a facility. These proposed changes would not meet the definition of “backfitting” in § 50.109 or affect issue finality as provided for in 10 CFR part 52.

I. Revision to Training Requirements in § 50.120

The proposed amendments would revise the timing for when the training requirements in § 50.120 must be met and require only the training and qualification of applicable categories of plant personnel for the facility. These proposed changes would be nonmandatory relaxations because facility licensees that would not have loaded fuel ( i.e., would not have previously operated) could establish, implement, and maintain a training program more than 18 months prior to initial fuel load and could provide for a training program for all the currently listed categories of plant personnel. Therefore, the proposed changes to the training requirements in § 50.120 would not meet the definition of “backfitting” in § 50.109 or affect issue finality as provided for in 10 CFR part 52.

J. Elimination of Wait Times for Re-Applications in § 55.35

The proposed amendments would eliminate waiting times between re-applications for operator licenses. This proposed changes would be nonmandatory relaxations. The applicant and facility licensee could wait to submit re-applications consistent with the current wait times. Therefore, the proposed changes to the re-application requirements in § 55.35 would not meet the definition of “backfitting” in § 50.109 or affect issue finality as provided for in 10 CFR part 52.

K. Operating Tests

The proposed amendments would allow facility licensees to proctor and grade the operating tests required by § 55.45. This proposed change would provide voluntary alternatives because licensees would not be required to proctor and grade operating tests. Furthermore, proposed changes to operating test administration requirements would not modify procedures or organizations required to operate a facility. Therefore, the proposed changes to the operating tests would not meet the definition of “backfitting” in § 50.109 or affect issue finality as provided for in 10 CFR part 52.

L. Plant Walkthrough Requirements in 10 CFR 55.45

A new provision would be added to § 55.45(b) to allow for suitable alternatives to the plant walkthrough portion of an operating test when a facility is under construction. This proposed change would provide licensees with a voluntary alternative because licensees would not be required to provide alternatives to a plant walkthrough. Editorial changes to § 55.45(b) would also be made that would not modify the current requirements in § 55.45(b). Furthermore, changes to plant walkthrough requirements would not modify procedures or organizations required to operate a facility. Therefore, the proposed changes to the operating tests would not meet the definition of “backfitting” in § 50.109 or affect issue finality as provided for in 10 CFR part 52.

M. Examination Waiver Requirements

The proposed amendments would make editorial changes and add criteria to § 55.47 for the NRC to grant a waiver of examination and operating test requirements at multiunit sites. The proposed changes to the waiver requirements would not meet the definition of “backfitting” in § 50.109 or affect issue finality as provided for in 10 CFR part 52 because these proposed changes would apply to the NRC and would not be imposed on licensees, applicants, or other entities.

N. Removing Obsolete Information From § 55.5

The proposed amendments would remove examples of methods to submit information to the NRC electronically but continue to permit electronic submissions where practicable. Additionally, the proposed amendments would remove NRC contact information from this section and instead refer to the contact information in appendix D to 10 CFR part 20. These changes would not modify procedures or organizations required to operate a facility. Therefore, the proposed changes to § 55.5 would not meet the definition of “backfitting” in § 50.109 or affect issue finality as provided for in 10 CFR part 52.

O. Elimination of License Renewal for 10 CFR Part 55 Operator's Licenses

The proposed amendments would revise the requirements in § 55.55 such that an expiration date would no longer be included on operator licenses issued in the future. This change would not modify the expiration date of current operator licenses. Additionally, the operator license renewal requirements would be eliminated as they would no longer be necessary when the operator license does not expire. The proposed changes to the operator license expiration provisions and the elimination of the license renewal ( printed page 60750) provisions would not modify procedures or organizations required to operate a facility. Therefore, the proposed changes to § 55.5 would not meet the definition of “backfitting” in § 50.109 or affect issue finality as provided for in 10 CFR part 52.

P. Generally Licensed Reactor Operators for Self-Reliant Mitigation Facilities

There are currently no licensed SRMFs. The proposed changes would provide a new option for prospective applicants, so it would not result in the modification of or addition to the design or organization necessary to operate any existing facilities. Therefore, the proposed changes to § 55.74 through 79 would not meet the definition of “backfitting” in § 50.109 or affect issue finality as provided for in 10 CFR part 52. The proposed changes to §§ 50.34(b)(6) and 52.79(a)(26) would also not meet the definition of “backfitting” in § 50.109 or affect issue finality as provided for in part 52 because there are currently no licensed SRMFs.

10 CFR Part 75—IAEA Notifications

The proposed amendments would delete § 75.13(c), which is bounded by the requirements of § 75.13(b)(3) and therefore redundant. Therefore, the proposed elimination of § 75.13(c) would not meet the definition of “backfitting” in §§ 50.109, 70.76, and 72.62 or affect issue finality as provided for in 10 CFR part 52 because the proposed change would not impose new or modified requirements on any license or approvals within the scope of these regulations.

10 CFR Parts 50, 52, 53, 70, and 72—Recordkeeping and Reporting Requirements

The proposed amendments would eliminate several recordkeeping and reporting requirements. The elimination of these requirements would be a nonmandatory relaxation, meaning that licensees or other entities subject to these requirements could continue to maintain records and submit reports to the NRC consistent with the current requirements. Therefore, the proposed elimination of recordkeeping and reporting requirements would not meet the definition of “backfitting” in §§ 50.109, 70.76, and 72.62 or affect issue finality as provided for in 10 CFR part 52.

The proposed amendments to § 50.91 would change certain communication requirements that apply only to the NRC and its staff during the review of license amendment requests under 10 CFR part 50. Therefore, the proposed amendments to § 50.91 do not meet the definitions of “backfitting” in § 50.109 or affect issue finality for NRC approvals issued under 10 CFR part 52.

The proposed amendment to § 70.32(g) would change the reporting requirement timeframe from 60 days to 4 months. The increased timeframe allowed in these requirements would be a nonmandatory relaxation, meaning that licensees or other entities subject to these requirements could continue to maintain records and submit reports to the NRC consistent with the current requirements. Therefore, the proposed change to the reporting requirements in § 70.32(g) would not meet the definition of “backfitting” in § 70.76.

The proposed amendment to § 70.59 would change the reporting requirement from “within 60 days after January 1 and July 1 of each year,” to “within 60 days after January 1 of each year,” for each licensee authorized to possess and use special nuclear material in a spent fuel reprocessing facility. The increased timeframe allowed in these requirements would be a nonmandatory relaxation, meaning that licensees or other entities subject to these requirements could continue to maintain records and submit reports to the NRC consistent with the current requirements. For other facilities, the report would only be required should the annual radiation doses to the public, resulting from the release result in an estimated total effective dose equivalent, or total effective dose, as applicable, of more than 25 mrem per year, or if quantities of radioactive materials released during the reporting periods are significantly above the licensee's design objectives previously reviewed as part of the licensing action. The change in these requirements would be a nonmandatory relaxation, meaning that licensees or other entities subject to these requirements could continue to maintain records and submit reports to the NRC consistent with the current requirements. Therefore, the proposed changes to the recordkeeping and reporting requirements in § 70.59 would not meet the definition of “backfitting” in § 70.76.

Guidance

The NRC is issuing revisions to four NUREG-series documents for public comment with this proposed rule. NUREG-1022, “Event Reporting Guidelines: 10 CFR 50.72 and 10 CFR 50.73,” provides guidelines that the NRC staff considers acceptable for reactor licensees to use in meeting the requirements of §§ 50.72 and 50.73. If issued in final form, this revision of NUREG-1022 would not be imposed on licensees and would not require the modification of procedures or organizations required to operate a reactor facility. Therefore, issuance of NUREG-1022 in final form would not meet the definition of “backfitting” in §§ 50.109, 70.76, and 72.62; constitute forward fitting, as that term is defined in MD 8.4; or affect issue finality as provided for in 10 CFR part 52.

The draft revision of NUREG-1409, “Backfitting Guidelines,” would provide guidance for the NRC staff to implement the NRC's backfitting, forward fitting, and issue finality provisions and policies. This draft NUREG would also provide guidance for external entities to raise backfitting, forward fitting, or issue finality concerns and for affected entities to submit formal appeals per MD 8.4. If issued in final form, this revision of NUREG-1409 would not be imposed on licensees and would not require the modification of procedures or organizations required to operate a reactor facility. Therefore, issuance of NUREG-1409 in final form would not meet the definition of “backfitting” in §§ 50.109, 70.76, and 72.62; constitute forward fitting, as that term is defined in MD 8.4; or affect issue finality as provided for in 10 CFR part 52.

NUREG-1021, “Operator Licensing Examination Standards for Power Reactors,” and NUREG-1478, “Operator Licensing Examiner Standards for Research and Test Reactors,” would be revised to support the implementation of the changes to operator licensing requirements in this proposed rule. Currently, power reactor facility licensees may prepare, proctor, and grade written examinations and may prepare operating tests for operators in accordance with § 55.40. If a power reactor facility licensee chooses this option, they must prepare the examinations and tests in accordance with the criteria in NUREG-1021 in effect 6 months before the examination date and receive written approval of the examinations and tests before they are administered. This proposed rule would allow power reactor licensees to also proctor and grade the operating tests and use other criteria approved by the NRC. In addition, these options would be made available to non-power reactor licensees who could use the criteria in NUREG-1478 or other criteria approved by the NRC. NUREG-1021 and NUREG-1478 would not be imposed on reactor licensees. Additionally, the examinations and tests are prospective in nature as they require prior NRC approval before administration. The issuance of these NUREG revisions in ( printed page 60751) final form would not modify procedures or organizations required to operate a reactor facility. Therefore, issuance of NUREG-1021 and NUREG-1478 in final form would not meet the definition of “backfitting” in §§ 50.109, 70.76, and 72.62; constitute forward fitting, as that term is defined in MD 8.4; or affect issue finality as provided for in 10 CFR part 52.

The NRC is issuing four draft regulatory guides that, if finalized, would provide guidance on the methods acceptable to the NRC for complying with aspects of this proposed rule. As discussed in the draft regulatory guides, applicants and licensees would not be required to comply with the positions set forth in the draft regulatory guides. Therefore, issuance of the draft regulatory guides in final form would not meet the definition of “backfitting” in §§ 50.109, 70.76, and 72.62; constitute forward fitting, as that term is defined in MD 8.4; or affect issue finality as provided for in 10 CFR part 52.

IX. Cumulative Effects of Regulation

The NRC seeks to minimize potential negative consequences resulting from the cumulative effects of regulation (CER). The NRC believes that the de-regulatory impacts of this rulemaking activity are unlikely to cause implementation challenges for stakeholders. In addition, during the pendency of this rulemaking, the NRC is deprioritizing issuance of regulatory actions that might influence the implementation date for the new rule requirements ( e.g., orders, generic communications, license amendment requests, and inspection findings of a generic nature).

To fully understand any potential CER implications that could result from this rulemaking, the NRC is asking the following questions. Response to these questions is voluntary and any input will be considered during development of the final rule.

1. The NRC is proposing an effective date that will be 30 days after the date of publication of a final rule. Does this provide sufficient time to implement the proposed requirements? Please provide a rationale for your response.

2. Are there unintended consequences related to this rulemaking and how should they be addressed? Please provide a rationale for your response.

3. Please comment on the NRC's cost and benefit estimates in the regulatory analysis that supports this proposed rule.

X. Plain Writing

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

XI. National Environmental Policy Act

The Commission has determined under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in subpart A of 10 CFR part 51, that this proposed rule, if adopted, would not be a major Federal action significantly affecting the quality of the human environment, and an environmental impact statement is not required. The basis for this determination is the implementation of this proposed rule described in this Federal Register notice would not have a significant environmental impact. The proposed regulatory changes are administrative in application or are matters of procedure, record keeping and reporting requirements, and provide equivalent levels of safety as existing requirements. This rulemaking would not materially change any regulatory requirements and would result in no new or different environmental effects

As explained in the environmental assessment, the proposed agency action ( i.e., rulemaking) will not have a significant effect on the quality of the human environment. Public comments on the environmental assessment may be submitted to the NRC as indicated under the ADDRESSES caption. The environmental assessment is listed under the “Availability of Documents” section.

XII. Paperwork Reduction Act

This proposed rule contains new or amended collections of information subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). This proposed rule has been submitted to the Office of Management and Budget (OMB) for review and approval of the information collections.

Type of submission: New.

The title of the information collection: Regulatory Enhancement for Reactor Licensing, Decommissioning, and Operational Oversight.

The form number if applicable: NRC Forms 361, 396, 398, and 536.

How often the collection is required or requested: Reporting frequencies vary. Some reports, such as defect notifications under 10 CFR part 21, must be submitted promptly following discovery (typically within 60 days, with initial notifications in some cases required within 2 days), while event notifications for nuclear power reactors under § 50.72 are required immediately or within specified hours depending on the event's significance. Periodic reports include annual effluent monitoring reports for spent fuel reprocessing facilities, 24-month FSAR updates for power reactor licensees, 5-year FSAR updates for non-power production or utilization facilities, and 4-month reporting for changes to safeguards contingency plans. Operator requalification and medical certifications are generally required every 2 years, and some collections are only triggered when certain thresholds are exceeded ( e.g., effluent reports if annual public dose exceeds 25 mrem/year). Records retention periods also vary: many records must be kept for the duration of the license, superseded records are typically retained for at least 3 years after a change, and certain procurement, evaluation, and notification records must be maintained for 10 or 15 years depending on the regulation.

Who will be required or asked to respond: Existing and future applicants and licensees under 10 CFR parts 21, 50, 52, 53, 55, 70, and 72.

An estimate of the number of annual responses:

10 CFR part 21: 34 (4 reporting responses + 30 recordkeepers)

10 CFR part 50: −794.6 (−821.1 reporting responses + 26.5 recordkeepers)

10 CFR part 50.55a: −9 (−9 reporting responses)

10 CFR part 52: −35.7 (−35.7 reporting responses)

10 CFR part 53: −11.3 (−16.3 reporting responses + 5 recordkeepers)

10 CFR part 55: 203.6 (202.6 reporting responses + 1 recordkeeper)

10 CFR part 70: −16 (−24 reporting responses + 8 recordkeepers)

10 CFR part 72: −102 (−118 reporting responses + 16 recordkeepers)

Form 361: −244.4 (−244.4 reporting responses)

Form 396: −1,197 (−598.5 reporting responses + −598.5 recordkeepers)

Form 398: −402 (−402 reporting responses)

Form 536: 68 (68 reporting responses)

The estimated number of annual respondents:

10 CFR part 21: 30 respondents

10 CFR part 50: 119 respondents

10 CFR part 50.55a: −9 respondents ( printed page 60752)

10 CFR part 52: 9.33 respondents

10 CFR part 53: 5 respondents

10 CFR part 55: 67 respondents

10 CFR part 70: 8 respondents

10 CFR part 72: 16 respondents

Form 361: −94 respondents

Form 396: −133 respondents

Form 398: −402 respondents

Form 536: 68 respondents

An estimate of the total number of hours needed annually to comply with the information collection requirement or request:

10 CFR part 21: 3,160 (160 reporting + 3,000 recordkeeping)

10 CFR part 50: −104,514.9 (−104,465.7 reporting + −49.2 recordkeeping)

10 CFR part 50.55a: −225 (−225 reporting)

10 CFR part 52: −70 (−70 reporting)

10 CFR part 53: −579.6 (−79.6 reporting + −500 recordkeeping)

10 CFR part 55: 2,599 (993 reporting + 1,606 recordkeeping)

10 CFR part 70: −64 (−3,360 reporting + 3,296 recordkeeping)

10 CFR part 72: −262 (−2,072 reporting + 1,810 recordkeeping)

Form 361: −122.2 (−122.2 reporting)

Form 396: −748.1 (−598.5 reporting + −149.6 recordkeeping)

Form 398: −1,005 (−1,005 reporting)

Form 536: 5.7 (5.7 reporting)

Abstract: The NRC is proposing to amend its regulations to provide regulatory enhancements for reactor licensing, decommissioning, and operational oversight. The NRC identified potential changes across various parts to 10 CFR, as well as NRC forms. This effort is consistent with, and implements direction in, the recently issued E.O. 14300.

This proposed rule covers a wide range of topics, including the following areas that would result in new or revised changes in recordkeeping and reporting requirements:

a. Part 21 Clarifications (parts 21, 50, and 53).

b. Reporting Requirements for Nonemergency Events for Nuclear Power Plants (part 50).

c. Decommissioning Licensing Enhancements (parts 20, 50, 52, and 53).

d. Financial Qualifications and Decommissioning Trust Fund Use (parts 50, 52, and 53).

e. Probabilistic Risk Assessment Requirements (part 50).

f. Notification of Initial Fuel Load (parts 52 and 53).

g. Requirements for Evaluation of Generic Issues and Operating Experience (part 52).

h. Standard Design Approvals (parts 50, 52, and 53).

i. Updates to the Operator Licensing Programs (parts 50, 53, and 55).

j. Recordkeeping and Reporting Requirements (parts 50, 52, 53, 70, and 72).

This supporting statement includes burden associated with revised information collection in parts 21, 50, 52, 53, 55, 70, and 72, as well as NRC Forms 361, 396, 398, and 536.

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

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

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

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

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

A copy of the OMB clearance package and proposed rule are available in the “Availability of Documents” section of this document or may be viewed free of charge by contacting the NRC's Public Document Room reference staff at 1-800-397-4209, at 301-415-4737, or by email to PDR.Resource@nrc.gov. You may obtain information and comment on submissions related to the OMB clearance package by searching on www.regulations.gov under Docket ID NRC-2025-1138.

You may submit comments on any aspect of these proposed information collection(s), including suggestions for reducing the burden and on the above issues, by the following method:

Federal rulemaking Website: Go to www.regulations.gov and search for Docket ID NRC-2025-1138.

Submit comments by October 26, 2026.

Public Protection Notification

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

XIII. Executive Orders

The following are Executive orders that are related to this proposed rule:

A. Executive Order 12866: Regulatory Planning and Review (as Amended by Executive Order 14215, Ensuring Accountability for All Agencies)

The Office of Information and Regulatory Affairs (OIRA) has determined that this proposed rule is a significant regulatory action. Accordingly, the NRC submitted this proposed rule to OIRA for review. The NRC is required to conduct an economic analysis in accordance with section 6(a)(3)(B) of E.O. 12866. More information can be found in Section VII of this document, “Regulatory Analysis.”

B. Executive Order 14154: Unleashing American Energy

The NRC has examined this proposed rule and has determined that it is consistent with the policies and directives outlined in E.O. 14154.

C. Executive Order 14192: Unleashing Prosperity Through Deregulation

This action is a deregulatory action as defined by E.O. 14192. Details on the estimated costs of this proposed rule can be found in Section VII of this document, “Regulatory Analysis.”

D. Executive Order 14267: Reducing Anti-Competitive Regulatory Barriers

E.O. 14267 requires the NRC to identify anti-competitive regulations for recission or modification.

The NRC identified § 21.3, specifically the definition for “ dedicating entity,” because this regulation has an anticompetitive effect by limiting the pool of potential vendors for SSCs. Specifically, this proposed rule would eliminate unnecessary restrictions ( i.e., removing the reference to “nuclear power plants licensed under Part 50” and deleting the description of who can perform the dedication) that would limit vendor participation in supplying SSCs.

The NRC identified § 50.33 because this regulation has an anticompetitive effect by creating barriers to entry that are not consistent for all applicants. The financial qualification requirements in 10 CFR part 53 contain a less restrictive qualification standard of “appears to be financially qualified,” and § 50.33 would be updated to include this less restrictive standard.

The NRC identified § 50.54 because this regulation has an anticompetitive effect by raising barriers to entry. Specifically, paragraphs (m) and (j) of this section are being proposed to update the operator licensing requirements to include generally licensed reactor operators in addition to operator and senior operator. ( printed page 60753)

The NRC identified the compliance backfitting provisions in 10 CFR parts 50, 52, 53, 70, and 72 as having an anticompetitive effect. Specifically, due to the lack of specific criteria in the compliance backfitting provisions, the compliance justifications are susceptible to less objective application and, therefore, are less predictable and may create some regulatory uncertainty for entities when making long-term financial decisions regarding, for example, new facility construction or major upgrades. Such uncertainty may create a barrier to competition and result in higher regulatory burdens for new market entrants than imposed on legacy participants. Therefore, in response to E.O.14267, the NRC is proposing to eliminate the compliance backfitting provisions from its rules.

The NRC identified § 55.4 because the definition of “ physician” in this regulation creates a barrier to market participation by limiting the pool of eligible employees to work in this role. The NRC has included a new definition, “ licensed medical examiner,” and has referenced this new definition as an alternative to physician throughout 10 CFR parts 53 and 55.

The proposed modifications of the regulations support the objectives of E.O. 14267 by removing regulatory requirements that could create “unnecessary barriers to entry for new market entrants” or other applicable reason from E.O. 14267.

E. Executive Order 14270: Zero-Based Regulatory Budgeting To Unleash American Energy

E.O. 14270, “Zero-Based Regulatory Budgeting to Unleash American Energy,” requires the NRC to insert a conditional sunset date into all new or amended NRC regulations provided the regulations are (1) promulgated under the AEA, the Energy Reorganization Act of 1974, as amended (ERA), or the Nuclear Waste Policy Act of 1982, as amended (NWPA); (2) not statutorily required; and (3) not part of the NRC's permitting regime. The NRC determined that the regulatory changes proposed in this rule are part of the NRC's regulatory permitting scheme authorized by the AEA, ERA, or NWPA. Therefore, the NRC views this rulemaking to be outside the scope of E.O. 14270 and did not insert conditional sunset dates for the regulatory changes in this proposed rule.

XIV. Compatibility of Agreement State Regulations

On the basis of the “Agreement State Program Policy Statement” approved by the Commission on October 2, 2017, and published in the Federal Register (82 FR 46840; October 6, 2017, and 82 FR 48535; October 18, 2017), NRC program elements can be placed into six categories (A, B, C, D, NRC, or health and safety) to form the basis for evaluating and classifying the program elements. Under the Policy Statement, Compatibility Category NRC are those program elements that address areas of regulation that cannot be relinquished to the Agreement States under the AEA, or provisions of 10 CFR. The NRC maintains regulatory authority over these program elements and the Agreement States must not adopt these NRC program elements.

Although this rulemaking proposes changes to 10 CFR parts 20 and 70, the specific changes to those regulations, or sections of regulations, are designated Category NRC and, therefore, this rulemaking is not a matter of compatibility. However, an Agreement State may inform its licensees of these NRC requirements through a mechanism under the State's administrative procedure laws, as long as the State adopts these provisions solely for the purposes of notification, and does not exercise any regulatory authority as a result.

For completeness, a table showing the relevant regulations and the categories is below.

XV. Availability of Guidance

The NRC is issuing for comment eight revised draft guidance documents to support the implementation of the proposed requirements in this rulemaking. You may obtain information and comment submissions related to the draft guidance by searching on www.regulations.gov under Docket ID NRC-2025-1138. You may submit comments on these draft guidance documents by the methods outlined in the ADDRESSES section of this document.

1. Draft NUREG-1021, Revision 13, Operator Licensing Examination Standards for Power Reactors, Draft Report

2. Draft NUREG-1022, Revision 4, Event Report Guidelines: 10 CFR 50.72 and 10 CFR 50.73, Draft Report

3. Draft NUREG-1409, Revision 1, Backfitting Guidelines

4. Draft NUREG-1478, Revision 3, Operator Licensing Examiner Standards for Research and Test Reactors

5. DG-1469 (Draft RG 1.134, Revision 5), Medical Assessment of Licensed Operators or Applicants for Operator Licenses at Nuclear Power Plants

6. DG-1470 (Draft RG 1.149, Revision 5), Nuclear Power Plant Simulation Facilities For Use in Operator Training, License Examinations, and Applicant Experience Requirements ^[5]

( printed page 60754)

7. DG-1471 (Draft RG 1.208, Revision 2), A Performance-Based Approach to Define the Site-Specific Earthquake Ground Motion

8. DG-1476 (Draft RG 1.159 Revision 4), Assuring the Availability of Funds for Decommissioning Nuclear Reactors

XVI. Availability of Documents

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

( printed page 60758)

The NRC may post materials related to this document, including public comments, on the Federal rulemaking website at www.regulations.gov under Docket ID NRC-2025-1138. In addition, the Federal rulemaking website allows members of the public to receive alerts when changes or additions occur in a docket folder. To subscribe: (1) navigate to the docket folder (NRC-2025-1138); (2) click the “Subscribe” button; and (3) enter an email address and click on the “Subscribe” button.

For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended; the Energy Reorganization Act of 1974, as amended; and 5 U.S.C. 552 and 553, the NRC is proposing to amend 10 CFR parts 20, 21, 50, 52, 53, 55, 70, 72, and 75.

1. The authority citation for part 20 continues to read as follows:

Authority: Atomic Energy Act of 1954, secs. 11, 53, 63, 65, 81, 103, 104, 161, 170H, 182, 186, 223, 234, 274, 1701 (42 U.S.C. 2014, 2073, 2093, 2095, 2111, 2133, 2134, 2201, 2210h, 2232, 2236, 2273, 2282, 2021, 2297f); Energy Reorganization Act of 1974, secs. 201, 202 (42 U.S.C. 5841, 5842); Low-Level Radioactive Waste Policy Amendments Act of 1985, sec. 2 (42 U.S.C. 2021b); 44 U.S.C. 3504 note.

2. In § 20.1401, in paragraphs (a) and (c), remove the references “§ 50.83 or § 53.1080” and add in its place the references “§ 50.83, § 52.111, or § 53.1080”.

3. The authority citation for part 21 continues to read as follows:

Authority: Atomic Energy Act of 1954, secs. 53, 63, 81, 103, 104, 161, 223, 234, 1701 (42 U.S.C. 2073, 2093, 2111, 2133, 2134, 2201, 2273, 2282, 2297f); Energy Reorganization Act of 1974, secs. 201, 206 (42 U.S.C. 5841, 5846); Nuclear Waste Policy Act of 1982, secs. 135, 141 (42 U.S.C. 10155, 10161); 44 U.S.C. 3504 note.

4. In § 21.2:

a. Add new paragraph (a)(5);

b. Remove and reserve paragraph (b); and

c. Remove paragraph (e).

The addition reads as follows:

5. In § 21.3:

a. In the definition for “ Basic component”, revise paragraph (1)(i) introductory text, redesignate paragraph (ii) as paragraph (iii), add new paragraph (ii), add the phrase “or appendix T to part 50” to newly redesignated paragraph (1)(iii), remove the reference “70,” in paragraph (3), and add new paragraphs (5) through (7);

b. In the definition for “ Commercial grade item”, add the reference “52,” in numerical order in paragraph (1) and remove the reference “30, ” from paragraph (2); and

c. Revise the definitions for “ Dedicating entity”, “ Dedication”, and “ Notification”.

The revisions read as follows:

6. Revise and republish § 21.21 to read as follows:

7. In § 21.51:

a. Redesignate paragraph (b) to paragraph (c); and

b. Add new paragraph (b).

The addition reads as follows.

8. The authority citation for part 50 continues to read as follows:

Authority: Atomic Energy Act of 1954, secs. 11, 101, 102, 103, 104, 105, 108, 122, 147, 149, 161, 181, 182, 183, 184, 185, 186, 187, 189, 223, 234 (42 U.S.C. 2014, 2131, 2132, 2133, 2134, 2135, 2138, 2152, 2167, 2169, 2201, 2231, 2232, 2233, 2234, 2235, 2236, 2237, 2239, 2273, 2282); Energy Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); Nuclear Waste Policy Act of 1982, sec. 306 (42 U.S.C. 10226); National Environmental Policy Act of 1969 (42 U.S.C. 4332); 44 U.S.C. 3504 note; ADVANCE Act of 2024, sec. 301 (42 U.S.C. 2133 note).

9. In § 50.2:

a. Add in alphabetical order the definitions for “ Generally licensed reactor operator” and “ Self-reliant-mitigation facility”; and

b. Delete the definitions for “ Basic component”, “ Construction or constructing”, “ Defect”, “ Deviation”, “ Director”, “ Discovery”, “ Evaluation”, “ Procurement document”, “ Responsible officer”, and “ Substantial safety hazard”.

The additions read as follows:

10. In § 50.4, remove and reserve paragraphs (b)(5)(ii), (b)(6) and (b)(7)(i).

11. In § 50.8:

a. In paragraph (b), remove the references “50.55a,” and “50.74,”;

b. In paragraph (c)(3), remove the phrase “control numbers” and add in its place the phrase “control number”; and

c. Add new paragraph (d).

The addition reads as follows:

12. In § 50.33, revise paragraphs (f)(1) and (2) to read as follows:

13. In § 50.34:

a. Add new paragraphs (b)(6)(i)(A) and (B);

b. In paragraph (b)(8), remove the reference “§ 55.59” and add in its place the reference phrase “§ 55.59 or § 55.78, as applicable,”;

c. Revise paragraph (f) introductory text and paragraph (f)(1) introductory text;

d. Remove paragraph (f)(1)(xii);

e. Remove and reserve paragraph (f)(2)(i);

f. In paragraph (f)(2)(iv), remove the phrase “console”; and

g. Remove and reserve paragraphs (f)(2)(vi), (f)(2)(ix), (f)(2)(xxiv), (f)(2)(xxv), (f)(3)(v), and (f)(3)(vi).

The additions and revisions read as follows:

14. In § 50.36:

a. Revise paragraphs (c)(1)(i)(A), (c)(1)(ii), and (c)(2)(i); and

b. Remove paragraphs (c)(7) and (c)(8).

The revisions read as follows:

15. In § 50.46, paragraph (a)(3)(ii), remove the references “§§ 50.55(e), 50.72, and 50.73” and add in its place the references “part 21 of this chapter, § 50.72, and § 50.73”.

16. In § 50.54:

a. Revise and republish the § 50.54 introductory text;

b. Revise paragraphs (a)(3), (i), (i-1), (j), and (m) introductory text;

c. In paragraph (m)(2) introductory text, remove the date “, by January 1, 1984”;

d. Revise paragraph (m)(3);

e. In paragraph (o), remove the reference “52.110(a)(1)” and in its place add the reference “52.110(a)”;

f. In paragraph (p)(2) introductory text, remove the phrase “2 months” and in its place add the phrase “6 months”;

g. Revise paragraph (q)(5);

h. Remove and reserve paragraph (w)(3);

i. In paragraph (x), remove the phrase “contained in a license issued under this part” and in its place add the phrase “contained in a license issued under this part or part 52”;

j. Revise paragraph (y); and

k. Revise paragraph (ff).

The revisions read as follows:

17. In § 50.55:

a. Remove and reserve paragraph (e); and

b. Revise paragraphs (f)(3) and (f)(4)(i).

The revisions read as follows:

18. In § 50.55a:

a. Remove and reserve paragraphs (b)(2)(xliii) and (b)(2)(xlviii); and

b. Remove paragraph (f)(7).

19. In § 50.59:

a. Revise paragraphs (b), (c)(1) introductory text, (c)(1)(i), (c)(2) introductory text, (c)(3), and (d)(3); and

b. Remove and reserve paragraph (d)(2).

The revisions read as follows:

20. In § 50.69:

a. Revise and republish the introductory text to paragraph (b)(1); and

b. Remove and reserve paragraph (b)(1)(iv).

The revision reads as follows:

21. In § 50.71:

a. Revise paragraphs (a), (e) introductory text, (e)(3)(i), (iii), and (iv), and (e)(4)(i) and (ii);

b. Remove and reserve paragraphs (b) and (e)(1) and (2);

c. In paragraph (e)(5), remove the word “replacement” and in its place add the word “update”;

d. In paragraph (f), remove the word “submittal” and in its place add the word “update”; and

e. In paragraph (h)(2), remove the reference “§ 52.110(a)” and in its place add the reference “§ 52.110(a)(1)”;

The revisions read as follows:

22. In § 50.72:

a. In paragraphs (a)(1) introductory text, (a)(2) through (5), (b)(1), (b)(2) introductory text, (b)(3), and (c) introductory text, remove the phrase “shall” and add in its place the phrase “must”;

b. In paragraphs (a)(1) introductory text and (c)(3), remove the phrase “NRC” and add in its place the phrase “NRC Headquarters”;

c. In paragraph (c) introductory text, remove the phrase “the telephone”;

d. Add new paragraph (a)(6);

e. Remove and reserve paragraphs (b)(2)(iv)(A), (b)(2)(xi), and (b)(3)(iv) through (xii);

f. In footnote 1, remove the number “50.36,”; and

g. Remove footnote 4.

The addition reads as follows:

23. In § 50.73:

a. In paragraph (b)(2)(ii)(F)( 3), remove the phrase “ www.archives.gov/​federal_​register/​code_​of_​federal_​regulations/​ibr_​locations.html.” and add in its places the phrase “ www.archives.gov/​federal_​register/​code_​of_​federal_​regulations/​ibr_​locations.html.”; and

b. Remove and reserve paragraph (c).

24. Remove and reserve § 50.74.

25. In § 50.75, revise paragraphs (h)(1)(iv) and (h)(2) as follows:

26. Revise and republish § 50.82 to read as follows:

27. In § 50.83, revise paragraphs (a) introductory text and (f) to read as follows:

28. In § 50.91:

a. Revise paragraphs (a)(6)(i)(B) and (a)(6)(iii); and

b. In paragraphs (b)(3) and (4), remove the word “telephone” and in its place add the word “contact”.

The revisions read as follows:

29. Revise and republish § 50.109 to read as follows:

30. § 50.120:

a. In paragraph (b)(1)(i), remove the phrase “, by 18 months prior to fuel load,”;

b. Revise paragraph (b)(1)(ii) and (b)(3); and

c. In paragraph (b)(2) introductory text, remove the phrase “plant personnel” and in its place add the phrase “plant personnel, as applicable to the facility”.

The revisions read as follow:

31. Revise and republish appendix S to part 50.

Appendix S to Part 50—Earthquake Engineering Criteria for Nuclear Power Plants

General Information

This appendix applies to applicants for a construction permit or operating license under part 50, or a design certification, combined license, design approval, or manufacturing license under part 52 of this chapter. Paragraphs IV.a.1.i, IV.a.1.ii, IV.4.b, and IV.4.c of this appendix apply to applicants for an early site permit under part 52.

I. Introduction

(a) Each applicant for a construction permit, operating license, design certification, combined license, design approval, or manufacturing license is required by §§ 50.34(a)(12), 50.34(b)(10), or 10 CFR 52.47, 52.79, 52.137, or 52.157, and General Design Criterion 2 of appendix A to this part, to design nuclear power plant structures, systems, and components important to safety to withstand the effects of natural phenomena, such as earthquakes, without loss of capability to perform their safety functions. Also, as specified in § 50.54(ff), nuclear power plants that have implemented the earthquake engineering criteria described herein must shut down if the criteria in paragraph IV(a)(3) of this appendix are exceeded.

(b) These criteria implement General Design Criterion 2 insofar as it requires structures, systems, and components important to safety to withstand the effects of earthquakes.

II. Scope

The evaluations described in this appendix are within the scope of investigations permitted by § 50.10(a).

III. Definitions

As used in these criteria:

Combined license means a combined construction permit and operating license with conditions for a nuclear power facility issued under subpart C of part 52 of this chapter.

Design approval means an NRC staff approval, issued under subpart E of part 52 of this chapter, of a final standard design for a nuclear power reactor of the type described in 10 CFR 50.22.

Design-basis ground motions (DBGMs) are the vibratory ground motions for which certain structures, systems, and components must be designed to remain functional. This term is used in the place of the safe-shutdown earthquake ground motion when there are multiple vibratory ground motions used for structures, systems, and components at the site, equivalent to defining multiple safe-shutdown earthquake ground motions.

Design certification means a Commission approval, issued under subpart B of part 52 of this chapter, of a standard design for a nuclear power facility.

Manufacturing license means a license, issued under subpart F of part 52 of this chapter, authorizing the manufacture of nuclear power reactors but not their installation into facilities located at the sites on which the facilities are to be operated.

Operating basis earthquake ground motion (OBE) is the vibratory ground motion for which those features of the nuclear power plant necessary for continued operation without undue risk to the health and safety of the public are designed to remain functional. The OBE is only associated with plant shutdown and inspection unless specifically selected by the applicant as a design input.

Response spectrum is a plot of the maximum responses (acceleration, velocity, or displacement) of idealized single-degree-of-freedom oscillators as a function of the natural frequencies of the oscillators for a given damping value. The response spectrum is calculated for a specified vibratory motion input at the oscillators' supports.

Safe-shutdown earthquake ground motion (SSE) is the vibratory ground motion for which certain structures, systems, and components must be designed to remain functional.

Structures, systems, and components (SSCs) required to withstand the effects of the SSE, DBGMs, or surface deformation are determined based on their risk-significance and the safety functions that they are required to perform during and after the vibratory ground motions. These generally include SSCs necessary to assure:

(1) The integrity of the reactor coolant boundary;

(2) The capability to shut down the reactor and maintain it in a safe-shutdown condition; or

(3) The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guideline exposures of § 50.34(a)(1).

Surface deformation is distortion of geologic strata at or near the ground surface by the processes of folding or faulting as a result of various earth forces. Tectonic surface deformation is associated with earthquake processes.

IV. Application to Engineering Design

The following are pursuant to the seismic and geologic design basis requirements of part 100 of this chapter:

(a) Vibratory Ground Motions.

(1) Safe Shutdown Earthquake and Design-Basis Ground Motions.

(i) The SSE or DBGMs must be derived from the site ground motion response spectra developed in accordance with part 100 of this chapter at the free ground surface and must take into consideration the risk-significance of SSCs and their safety functions. In view of the limited data available on vibratory ground motions of strong earthquakes, it is acceptable that the design response spectra be smoothed spectra. In addition, the horizontal component of the SSE or DBGMs in the free-field at the foundation level of the structures must be an appropriate response spectrum that is determined based on the risk-significance of SSCs and their safety functions.

(ii) The nuclear power plant must be designed so that, if the SSE or DBGMs occur(s), certain SSCs will remain functional and within applicable stress, strain, and deformation limits. In addition to seismic loads, applicable concurrent normal operating, functional, and accident-induced loads must be taken into account in the design of these certain SSCs. The design of the nuclear power plant must also take into account the possible effects of the SSE or DBGMs on the facility foundations by ground disruption, such as fissuring, lateral spreads, differential settlement, liquefaction, and landsliding.

(iii) The required safety functions of SSCs must be assured during and after the vibratory ground motion associated with the SSE or DBGMs through design, testing, or qualification methods.

(iv) The evaluation should consider, if applicable, soil-structure interaction effects and the expected duration of vibratory motion. It is permissible to design for inelastic behavior in some of these certain SSCs during the SSE or DBGMs and under the postulated concurrent loads, provided the necessary safety functions are maintained.

(2) Operating Basis Earthquake Ground Motion.

(i) The OBE must be characterized by response spectra. The value of the OBE must be set to one of the following choices:

(A) One-third or less of the SSE or minimum DBGM design response spectra. The requirements associated with this OBE in paragraph (a)(2)(i)(B)( I) can be satisfied without the applicant performing explicit response or design analyses, or

(B) A value greater than one-third of the SSE or minimum DBGM design response spectra. Analysis and design must be ( printed page 60771) performed to demonstrate that the requirements associated with this OBE in Paragraph (a)(2)(i)(B)( I) are satisfied. The design must take into account soil-structure interaction effects and the duration of vibratory ground motion.

( I) When subjected to the effects of the OBE in combination with normal operating loads, all SSCs of the nuclear power plant necessary for continued operation without undue risk to the health and safety of the public must remain functional and within applicable stress, strain, and deformation limits.

(3) Required Plant Shutdown. If vibratory ground motion exceeding that of the OBE or if significant plant damage occurs, the licensee must shut down the nuclear power plant. If SSCs necessary for the safe shutdown of the nuclear power plant are not available after the occurrence of the OBE, the licensee must consult with the Commission and must propose a plan for the timely, safe shutdown of the nuclear power plant. Prior to resuming operations, the licensee must demonstrate to the Commission that those features necessary for continued operation without undue risk to the health and safety of the public or necessary to maintain the licensing basis were either not functionally damaged or have been repaired.

(4) Required Seismic Instrumentation. Suitable instrumentation must be provided so that the seismic response of nuclear power plant features important to safety can be evaluated promptly after an earthquake.

(b) Surface Deformation. The potential for surface deformation must be taken into account in the design of the nuclear power plant by providing reasonable assurance that in the event of deformation, certain SSCs will remain functional. In addition to surface deformation induced loads, the design of safety features must take into account seismic loads and applicable concurrent functional and accident-induced loads. The design provisions for surface deformation must be based on its postulated occurrence in any direction and azimuth and under any part of the nuclear power plant, unless evidence indicates this assumption is not appropriate, and must take into account the estimated rate at which the surface deformation may occur.

(c) Seismically Induced Floods and Water Waves and Other Design Conditions. Seismically induced floods and water waves from either locally or distantly generated seismic activity and other design conditions determined pursuant to part 100 of this chapter must be taken into account in the design of the nuclear power plant so as to prevent undue risk to the health and safety of the public.

32. The authority citation for part 52 continues to read as follows:

Authority: Atomic Energy Act of 1954, secs. 103, 104, 147, 149, 161, 181, 182, 183, 185, 186, 189, 223, 234 (42 U.S.C. 2133, 2134, 2167, 2169, 2201, 2231, 2232, 2233, 2235, 2236, 2239, 2273, 2282); Energy Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); 44 U.S.C. 3504 note.

33. In § 52.1, add in alphabetical order the definitions “ Generally licensed reactor operator” and “ Self-reliant-mitigation facility”.

34. In § 52.3:

a. In paragraph (a), remove the phrase “, for example, via Electronic Information Exchange, email, or CD-ROM”;

b. In paragraph (a), remove the phrase “NRC's Web site at www.nrc.gov/​site-help/​e-submittals.html; by email to MSHD.Resource@nrc.gov ” and in its place add the phrase “NRC's website at www.nrc.gov/​site-help/​e-submittals.html; by email to MSHD.Resource@nrc.gov ”;

c. Remove and reserve paragraphs (b)(5)(ii) and (b)(6); and

d. In paragraph (b)(9), remove the reference “§ 52.110(a)(1)” and add in its place the reference “§ 52.110(a)(2)”.

35. Revise § 52.18 to read as follows:

36. In § 52.31:

a. In paragraph (a)(2)(i), remove the phrase “security;” and add in its place the phrase “security; or”;

b. Remove paragraph (a)(2)(ii); and

c. Redesignate paragraph (a)(2)(iii) as paragraph (a)(2)(ii).

37. In § 52.39:

a. Remove paragraph (a)(1)(i); and

b. Redesignate paragraphs (a)(1)(ii) through (iv) as paragraphs (a)(1)(i) through (iii).

38. In § 52.47, remove and reserve paragraphs (a)(21) and (22).

39. In § 52.59:

a. In paragraph (b)(1), remove the phrase “and security;” and add in its place “and security; or”;

b. Remove paragraph (b)(2); and

c. Redesignate paragraph (b)(3) as paragraph (b)(2).

40. In § 52.63:

a. Revise paragraphs (a)(1) introductory text and (a)(4);

b. Remove paragraphs (a)(1)(i) and (a)(4)(i) and (ii);

c. Redesignate paragraphs (a)(1)(ii) through (vii) as paragraphs (a)(1)(i) through (vi);

d. In redesignated paragraph (a)(1)(ii), remove the phrase “maintains protection to” and in its place add the phrase “maintains reasonable assurance of adequate protection of”; and

e. In paragraph (a)(2)(i), remove the reference “§ 52.63(a)(1)(ii)” and in its place add the reference “52.63(a)(1)(i)”.

The revisions read as follows:

41. In § 52.79:

a. Remove and reserve paragraphs (a)(20) and (a)(37);

b. Add new paragraphs (a)(26)(i) and (ii); and

c. In (a)(34), remove the phrase “of this chapter” and add in its place the phrase “or § 55.78, as applicable, of this chapter”.

The additions read as follows:

42. In § 52.103, remove the first sentence of paragraph (a) and add in its place the sentences “The licensee shall notify the NRC of its scheduled date for initial loading of fuel no later than 270 days before the scheduled date. If the scheduled date for initial loading of fuel is subsequently revised, the licensee shall notify the NRC of the update to its schedule within 30 days of revising the schedule.”

43. In § 52.110:

a. Add new introductory text;

b. Revise and republish paragraph (d);

c. Revise paragraphs (e), (g), (i) introductory text, and (i)(3); and

d. Remove and reserve paragraphs (h)(2) and (h)(3).

The revisions and addition read as follows:

44. In part 52, subpart C, add new § 52.111 in sequential order. The addition reads as follows:

45. In § 52.137, remove and reserve paragraphs (a)(21) and (22).

46. In § 52.145:

a. Remove the phrase “An approved design” from paragraph (a) and add in its place the phrase “An approved design (as updated)”; and

b. Add new paragraphs (d), (e), and (f).

The additions read as follows:

47. Revise § 52.147 to read as follows:

48. In § 52.157, remove and reserve paragraphs (f)(28) and (29).

49. In § 52.171:

a. Revise paragraph (a)(1);

b. Add new paragraphs (a)(1)(i) and (ii); and

c. In paragraph (b)(2), remove the phrase “, and that the special circumstances outweigh any decrease in safety that may result from the reduction in standardization caused by the departure”.

The revision and additions read as follows:

50. In § 52.179:

a. In paragraph (b)(1), remove the phrase “security;” and add in its place the phrase “security; or”;

b. Remove paragraph (b)(2); and

c. Redesignate paragraph (b)(3) as (b)(2).

51. In appendix A to part 52, section VIII, “Processes for Changes and Departures,”

a. In B, “Tier 2 Information,” revise paragraph 3 introductory text; and

b. In B, “Tier 2 Information,” remove paragraphs 3.a. and 3.b. ( printed page 60774)

The revision reads as follows:

Appendix A to Part 52—Design Certification Rule for the U.S. Advanced Boiling Water Reactor

VIII. Processes for Changes and Departures

B. Tier 2 Information

3. The Commission may not require new requirements on Tier 2 information by plant-specific order, while this appendix is in effect under § 52.55 or § 52.61, unless a modification is necessary for reasonable assurance of adequate protection of the public health and safety or the common defense and security.

52. In appendix D to part 52, section VIII, “Processes for Changes and Departures,”

a. In B, “Tier 2 information,” revise paragraph 3 introductory text; and

b. In B, “Tier 2 information,” remove paragraphs 3.a. and 3.b.

The revision reads as follows:

Appendix D to Part 52—Design Certification Rule for the AP1000 Design

VIII. Processes for Changes and Departures

B. Tier 2 information

3. The Commission may not require new requirements on Tier 2 information by plant-specific order, while this appendix is in effect under § 52.55 or § 52.61, unless a modification is necessary for reasonable assurance of adequate protection of the public health and safety or the common defense and security.

53. In appendix E to part 52, section VIII, “Processes for Changes and Departures,”

a. In B, “Tier 2 information,” revise paragraph 3 introductory text; and

b. In B, “Tier 2 information,” remove paragraphs 3.a. and 3.b.

The revision reads as follows:

Appendix E to Part 52—Design Certification Rule for the ESBWR Design

VIII. Processes for Changes and Departures

B. Tier 2 information

3. The Commission may not require new requirements on Tier 2 information by plant-specific order, while this appendix is in effect under § 52.55 or § 52.61, unless a modification is necessary for reasonable assurance of adequate protection of the public health and safety or the common defense and security.

54. In appendix F to part 52, section VIII, “Processes for Changes and Departures,”

a. In B, “Tier 2 Information,” revise paragraph 3 introductory text; and

b. In B, “Tier 2 Information,” remove paragraphs 3.a. and 3.b.

The revision reads as follows:

Appendix F to Part 52—Design Certification Rule for the APR1400 Design

VIII. Processes for Changes and Departures

B. Tier 2 Information

3. The Commission may not require new requirements on Tier 2 information by plant-specific order, while this appendix is in effect under § 52.55 or § 52.61, unless a modification is necessary for reasonable assurance of adequate protection of the public health and safety or the common defense and security.

55. In appendix G to part 52, section VIII, “Processes for Changes and Departures,”

a. In B “Tier 2 Information,” revise paragraph 3 introductory text; and

b. In B, “Tier 2 Information,” remove paragraphs 3.a. and 3.b.

The revision reads as follows:

Appendix G to Part 52—Design Certification Rule for NuScale

VIII. Processes for Changes and Departures

B. Tier 2 Information

3. The Commission may not require new requirements on Tier 2 information by plant-specific order, while this appendix is in effect under § 52.55 or § 52.61, unless a modification is necessary for reasonable assurance of adequate protection of the public health and safety or the common defense and security.

56. The authority citation for part 53 continues to read as follows:

Authority: Atomic Energy Act of 1954, secs. 11, 101, 103, 108, 122, 147, 161, 181, 182, 183, 184, 185, 186, 187, 189, 223, 234 (42 U.S.C. 2014, 2131, 2132, 2133, 2134, 2135, 2138, 2152, 2167, 2169, 2201, 2231, 2232, 2233, 2234, 2235, 2236, 2237, 2239, 2273, 2282); Energy Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); Nuclear Waste Policy Act of 1982, sec. 306 (42 U.S.C. 10226); National Environmental Policy Act of 1969 (42 U.S.C. 4332); 44 U.S.C. 3504 note; Pub. L. 115-439, 132 Stat. 5571.

57. In § 53.020, add in alphabetical order the definitions “ Licensed medical examiner” and “ Physician”.

58. In 53.040:

a. In paragraph (b)(5)(i), remove the phrase “plan;” and add in its place the phrase “plan; and”; and

b. Remove and reserve paragraphs (b)(5)(ii), (b)(6), and (b)(7).

59. In § 53.120:

a. In paragraph (c), remove the phrase “control numbers” and in its place add the phrase “control number”;

b. In paragraph (c)(1) and (2), remove the phrase “and 53.795,”; and

c. In paragraph (c)(1), remove the phrase “53.780” and in its place add the phrase “and 53.780”.

60. Remove and reserve § 53.605.

61. In § 53.725:

a. In paragraph (c), remove the phrase “ Simulation facility means” and add in its place the phrase “ Simulation facility or simulator means”; and

b. In paragraphs (c)(2), (4), (7), and (11), remove the phrase “controls” and add in its place the phrase “the controls”.

62. Revise and republish § 53.726 to read as follows:

63. In § 53.730, revise and republish paragraph (g)(3) to read as follows:

64. In § 53.735, remove the phrase “generally licensed reactor operator license there,” and add in its place the phrase “generally licensed reactor operator license,”.

65. In § 53.765, revise paragraphs (a) and (b)(1) as follows:

66. Revise and republish § 53.770 to read as follows:

67. In § 53.775, revise paragraph (b)(2) to read as follows:

68. In § 53.780:

a. Revise paragraphs (c)(1)(i) and (ii), (c)(2)(ii)(C), (c)(3), and (f); and

b. In paragraph (d), remove the phrase “equitable and ”.

The revisions read as follows:

69. In § 53.785:

a. Revise paragraph (h); and

b. Add new paragraph (l).

The revision and addition read as follows:

70. Revise and republish § 53.795 and the section heading to read as follows:

71. In § 53.810, in paragraph (g), remove the phrase the “the Commission” and add in its place the phrase “the Commission via the facility licensee”.

72. In § 53.815:

a. In paragraph (a), remove the phrase “each licensee” and add in its place the phrase “each facility licensee”;

b. In paragraph (b)(3)(vi), remove the phrase “their use” and add in its place the phrase “its use”;

c. In paragraph (d), remove the phrase “equitable and ”; and

d. In paragraph (e)(2)(v), remove the phrase “with § 53.815(d)” and add in its place the phrase “with paragraph (d) of this section”.

73. In § 53.820, remove the phrase “individual basis” and add in its place the phrase “individual operator basis”.

74. In § 53.830, revise paragraph (b) as follows:

75. In § 53.1045:

a. Remove and reserve paragraph (a)(2); and

b. Revise paragraphs (b)(4) and (c).

The revision reads as follows:

76. § 53.1060:

a. Remove and reserve paragraph (e); and

b. In the introductory text of paragraphs (g) and (h), remove the reference “paragraph (e) of this section” and add in its place the reference “§ 53.1070(d)”.

77. In § 53.1070, revise paragraphs (d)(1) and (2), (e), (g), and (i)(3) to read as follows:

78. In § 53.1080, revise paragraph (f) to read as follows:

79. In § 53.1149, revise paragraph (a) to read as follows:

80. In § 53.1176: ( printed page 60778)

a. In paragraph (a)(2)(i), remove the phrase “security;” and add in its place the phrase “security; or”;

b. Remove paragraph (a)(2)(ii); and

c. Redesignate paragraph (a)(2)(iii) as paragraph (a)(2)(ii).

81. In § 53.1188:

a. Remove paragraph (a)(1)(i); and

b. Redesignate paragraphs (a)(1)(ii) through (iv) as paragraphs (a)(1)(i) through (iii).

82. In § 53.1218, revise paragraph (b) to read as follows:

83. In § 53.1221:

a. In paragraph (a), remove the phrase “An approved design” and add in its place the phrase “An approved design (as updated)”; and

b. Add new paragraphs (e), (f), and (g).

The additions read as follows:

84. In § 53.1257:

a. In paragraph (a)(1), remove the phrase “security;” and add in its place the phrase “security; or”

b. Remove paragraph (b)(2); and

c. Redesignate paragraph (b)(3) as paragraph (b)(2).

85. Revise and republish § 53.1263 to read as follows:

86. In § 53.1288:

a. Revise paragraph (a)(1); and

b. Add new paragraphs (a)(1)(i) and (ii) and (a)(4).

The revision and additions read as follows:

87. In § 53.1295:

a. In paragraph (b)(2)(i), remove the phrase “security;” and add in its place the phrase “security; or”;

b. Remove paragraph (b)(2)(ii); and

c. Redesignate paragraph (b)(2)(iii) as paragraph (b)(2)(ii).

88. In § 53.1452, revise paragraph (a) and footnote 1 to read as follows:

89. In § 53.1505, revise paragraph (a) to read as follows:

90. In § 53.1515, revise paragraphs (a)(5)(i)(B) and (a)(5)(iii) to read as follows:

91. In § 53.1545:

a. Revise paragraphs (a), (c), and (e); and

b. Remove and reserve paragraph (b).

The revision reads as follows:

92. Remove and reserve paragraph (c)(2).

93. In § 53.1560: ( printed page 60780)

a. In the last sentence of paragraph (a) introductory text, remove the phrase “The submittals must” and add in its place “The update must”; and

b. Remove and reserve paragraph (b).

94. In § 53.1565:

a. In paragraph (b) remove the phrase “submittal of”; and

b. Revise paragraphs (d)(1)(i), (d)(2), and (d)(3)(v).

The revision reads as follows:

95. Revise and republish § 53.1590 as follows:

96. In § 53.1640, remove and reserve paragraph (c).

97. In § 53.1720, remove and reserve paragraph (c).

98. The authority citation for part 55 continues to read as follows:

Authority: Atomic Energy Act of 1954, secs. 107, 161, 181, 182, 183, 186, 187, 223, 234 (42 U.S.C. 2137, 2201, 2231, 2232, 2233, 2236, 2237, 2273, 2282); Energy Reorganization Act of 1974, secs. 201, 202 (42 U.S.C. 5841, 5842); Nuclear Waste Policy Act of 1982, sec. 306 (42 U.S.C. 10226); 44 U.S.C. 3504 note.

99. In § 55.1 paragraph (c), remove the phrase “and renew”.

100. In § 55.3, remove the phrase “or a senior operator” and replace it with the phrase “or a senior operator or a generally licensed reactor operator”.

101. In § 55.4:

a. Add in alphabetical order the definitions for “ Generally licensed reactor operator”, “ Licensed medical examiner”, and “ Self-reliant-mitigation facility”;

b. Remove the definition for “ Plant referenced simulator”; and

c. Revise the definitions for “ Licensee”, “ Reference Plant”, and “ Simulation facility”.

The additions and revisions read as follows:

102. In § 55.5:

a. Revise paragraphs (a)(3) and (b)(2) introductory text;

b. In paragraph (b)(1), remove the phrase “and renewal”;

c. Remove paragraphs (b)(2)(i) through (iv); and

d. In paragraph (b)(3), remove the phrase “or license renewal”.

The revisions read as follows:

103. In § 55.8, revise paragraphs (b) and (c)(2) to read as follows:

104. Revise § 55.13 to read as follows:

105. Revise § 55.21 to read as follows:

106. In § 55.23:

a. In § 55.23 introductory text, remove the phrase “Website at www.nrc.gov” and add in its place the phrase “website at www.nrc.gov”; and

b. Revise paragraph (a).

The revision reads as follows:

107. Revise and republish § 55.25 to read as follows:

108. Revise and republish § 55.27 to read as follows:

109. In § 55.31:

a. In paragraph (a)(1), remove the phone number “(301) 415-7232” and add in its place the phone number “(301) 415-5877”;

b. In paragraph (a)(1), remove the phrase “Web site at www.nrc.gov” and add in its place the phrase “website at www.nrc.gov”;

c. In paragraph (a)(4), remove the phrase “Form NRC-” and add in its place the phrase “NRC Form”;

d. Revise paragraph (a)(5); and

e. In paragraph (a)(6), remove the phrase “Form NRC-” and add in its place the phrase “NRC Form”.

The revision reads as follows.

110. In § 55.33, revise paragraph (b) to read as follows:

111. In § 55.35:

a. Revise paragraph (a); and

b. In paragraph (b), remove the phrase “Form NRC-” and add in its place the phrase “NRC Form”.

The revision reads as follows:

112. Revise and republish § 55.40 as follows:

113. In 55.41, revise paragraph (b) introductory text to read as follows:

114. In § 55.43, revise paragraph (b) introductory text as follows:

115. In § 55.45, revise paragraph (a) introductory text and paragraph (b) to read as follows:

116. Revise and republish § 55.46 as follows:

117. Revise and republish § 55.47 to read as follows:

118. In § 55.49, remove the phrase “equitable and ”.

119. In § 55.53:

a. Revise paragraphs (e), (f)(2), and (h);

b. In paragraph (g), remove the phrase “Commission” and add in its place the phrase “Commission via the facility licensee”; and

c. Add new paragraph (m).

The revisions and addition read as follows:

120. Revise and republish § 55.55 to read as follows:

121. Remove and reserve § 55.57.

122. Revise and republish § 55.59 to read as follows:

123. In § 55.73:

a. In paragraph (a), remove the phrase “conspiracy of violate” and in its place add the phrase “conspiracy to violate”; and

b. In paragraph (b), remove the reference “and 55.73” and add in its place the references “55.73, 55.74, and 55.79”.

124. In part 55, add new subpart I to read as follows:

125. The authority citation for part 70 continues to read as follows:

Authority: Atomic Energy Act of 1954, secs. 51, 53, 57(d), 108, 122, 161, 182, 183, 184, 186, 187, 193, 223, 234, 274, 1701 (42 U.S.C. 2071, 2073, 2077(d), 2138, 2152, 2201, 2232, 2233, 2234, 2236, 2237, 2243, 2273, 2282, 2021, 2297f); Energy Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); Nuclear Waste Policy Act of 1982, secs. 135, 141 (42 U.S.C. 10155, 10161); 44 U.S.C. 3504 note.

126. In § 70.32 ( printed page 60789)

a. In paragraph (g), remove the time period “60 days” and add in its place the time period “four months”; and

b. Revise paragraph (i) to read as follows:

127. Revise and republish § 70.59 to read as follows:

128. In § 70.72, revise paragraph (d)(3) to read as follows:

129. In § 70.76:

a. In paragraph (a)(1), remove the phrase “may result from” and add in its place the phrase “would be required by”;

b. Remove paragraphs (a)(4)(i) and (ii);

c. Redesignate paragraphs (a)(4)(iii) and (iv) as paragraphs (a)(4)(i) and (ii); and

d. Revise paragraph (a)(7).

The revision reads as follows:

130. The authority citation for part 72 continues to read as follows:

Authority: Atomic Energy Act of 1954, secs. 51, 53, 57, 62, 63, 65, 69, 81, 161, 182, 183, 184, 186, 187, 189, 223, 234, 274 (42 U.S.C. 2071, 2073, 2077, 2092, 2093, 2095, 2099, 2111, 2201, 2210e, 2232, 2233, 2234, 2236, 2237, 2238, 2273, 2282, 2021); Energy Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); National Environmental Policy Act of 1969 (42 U.S.C. 4332); Nuclear Waste Policy Act of 1982, secs. 117(a), 132, 133, 134, 135, 137, 141, 145(g), 148, 218(a) (42 U.S.C. 10137(a), 10152, 10153, 10154, 10155, 10157, 10161, 10165(g), 10168, 10198(a)); 44 U.S.C. 3504 note.

Source: 53 FR 31658, Aug. 19, 1988, unless otherwise noted.

131. In § 72.44, revise paragraph (f) to read as follows:

132. In § 72.62, revise paragraph (b) to read as follows:

133. In § 72.70, revise and republish paragraph (c) to read as follows:

134. In § 72.75:

a. Remove and reserve paragraph (b)(2); and

b. Remove paragraph (c)(3).

135. In § 72.80, remove and reserve paragraph (b).

136. In § 72.248, revise and republish paragraph (c) to read as follows:

137. The authority citation for part 75 continues to read as follows:

Authority: Atomic Energy Act of 1954, secs. 53, 63, 103, 104, 122, 161, 223, 234, 1701 (42 U.S.C. 2073, 2093, 2133, 2134, 2152, 2201, 2273, 2282, 2297f); Energy Reorganization Act of 1974, sec. 201 (42 U.S.C. 5841); Nuclear Waste Policy Act of 1982, secs. 135, 141 (42 U.S.C. 10155, 10161); 44 U.S.C. 3504 note.

Section 75.4 also issued under Nuclear Waste Policy Act secs. 135 (42 U.S.C. 10155, 10161).

138. In § 75.13, remove and reserve paragraph (c).

91 FR 60702